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Sample records for safety demonstration analyses

  1. Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

    Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

    2005-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

  2. Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

    Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

    2004-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

  3. Safety demonstration analyses on criticality for severe accident during overland transport of fresh nuclear fuel

    Takahashi, Satoshi; Okuno, Hiroshi; Yamada, Kenji; Watanabe, Kouji; Nomura, Yasushi; Miyoshi, Yoshinori

    2005-01-01

    Criticality safety analysis was performed for transport packages of uranium dioxide powder or of fresh PWR fuel involved in a severe accident during overland transportation, and as a result, sub-criticality was confirmed against impact accident conditions such as loaded by a drop from high position to a concrete or asphalt surface, and fire accident conditions such as caused by collisions with an oil tank trailer carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside an unventilated tunnel. (author)

  4. Spacecraft Fire Safety Demonstration

    National Aeronautics and Space Administration — The objective of the Spacecraft Fire Safety Demonstration project is to develop and conduct large-scale fire safety experiments on an International Space Station...

  5. Safety analyses for NHR-200

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The NHR-200 is a commercial 200-MW District Heating Reactor developed in China. It is designed on the basis of design, construction and four-year operating experience of the 5MW Experimental Heating Reactor (NHR-5). It has special safety features which are briefly described in this paper. Accident classification and safety criteria are also explained. Some typical and serious accidents are studied theoretically, and their results are detailed in this paper. They demonstrate the excellent safety characteristics of HR-200. (author). 4 refs, 9 figs, 1 tab.

  6. Chapter No.4. Safety analyses

    2002-01-01

    In 2001 the activity in the field of safety analyses was focused on verification of the safety analyses reports for NPP V-2 Bohunice and NPP Mochovce concerning the new profiled fuel and probabilistic safety assessment study for NPP Mochovce. The calculation safety analyses were performed and expert reviews for the internal UJD needs were elaborated. An important part of work was performed also in solving of scientific and technical tasks appointed within bilateral projects of co-operation between UJD and its international partnership organisations as well as within international projects ordered and financed by the European Commission. All these activities served as an independent support for UJD in its deterministic and probabilistic safety assessment of nuclear installations. A special attention was paid to a review of probabilistic safety assessment study of level 1 for NPP Mochovce. The probabilistic safety analysis of NPP related to the full power operation was elaborated in the study and a contribution of the technical and operational improvements to the risk decreasing was quantified. A core damage frequency of the reactor was calculated and the dominant initiating events and accident sequences with the major contribution to the risk were determined. The target of the review was to determine the acceptance of the sources of input information, assumptions, models, data, analyses and obtained results, so that the probabilistic model could give a real picture of the NPP. The review of the study was performed in co-operation of UJD with the IAEA (IPSART mission) as well as with other external organisations, which were not involved in the elaboration of the reviewed document and probabilistic model of NPP. The review was made in accordance with the IAEA guidelines and methodical documents of UJD and US NRC. In the field of calculation safety analyses the UJD activity was focused on the analysis of an operational event, analyses of the selected accident scenarios

  7. Public requirement to demonstrate safety

    Green, P.

    1991-01-01

    To many working within Government or industry, public concern over the disposal of radioactive waste is misplaced and has arisen out of an irrational and unscientific fear of technology, or even science in general. Members of the public, it is argued, are concerned because they do not understand the size of the risk in question. From the industry's point of view, the risk arising from the disposal of radioactive waste is ''negligible when compared to other everyday risks of life. Furthermore, any public exposure that may arise, either soon after closure of a facility or in the far future would comply with internationally accepted safety standards. In this context, the continuing concern over disposal of radioactive waste is viewed as evidence of the irrational and unscientific attitude of the public. The assessment and regulation of risk from waste disposal therefore is presented as a purely scientific question. Some of these issues are examined and public concern is shown not to be irrational but to be based upon legitimate questions over current waste management policy. An important question is not just ''how safe is safe, but who decides and how?''. (Author)

  8. Response surface use in safety analyses

    Prosek, A.

    1999-01-01

    When thousands of complex computer code runs related to nuclear safety are needed for statistical analysis, the response surface is used to replace the computer code. The main purpose of the study was to develop and demonstrate a tool called optimal statistical estimator (OSE) intended for response surface generation of complex and non-linear phenomena. The performance of optimal statistical estimator was tested by the results of 59 different RELAP5/MOD3.2 code calculations of the small-break loss-of-coolant accident in a two loop pressurized water reactor. The results showed that OSE adequately predicted the response surface for the peak cladding temperature. Some good characteristic of the OSE like monotonic function between two neighbor points and independence on the number of output parameters suggest that OSE can be used for response surface generation of any safety or system parameter in the thermal-hydraulic safety analyses.(author)

  9. Evaluation of periodic safety status analyses

    Faber, C.; Staub, G.

    1997-01-01

    In order to carry out the evaluation of safety status analyses by the safety assessor within the periodical safety reviews of nuclear power plants safety goal oriented requirements have been formulated together with complementary evaluation criteria. Their application in an inter-disciplinary coopertion covering the subject areas involved facilitates a complete safety goal oriented assessment of the plant status. The procedure is outlined briefly by an example for the safety goal 'reactivity control' for BWRs. (orig.) [de

  10. Summary view on demonstration reactor safety

    Satoh, Kazuziro; Kotake, Shoji; Tsukui, Yutaka; Inagaki, Tatsutoshi; Miura, Masanori

    1991-01-01

    This work presents a summary view on safety design approaches for the demonstration fast breeder reactor (DFBR). The safety objective of DFBR is to be at lea as safe as a LWR. Major safety issues discussed in this paper are; reduction of sodium void reactivity worth, adoption of self-actuated mechanism in the backup shutdown system, use of the direct reactor auxiliary cooling system (DRACS), provision of the containment system. (author)

  11. Periodic safety analyses; Les essais periodiques

    Gouffon, A; Zermizoglou, R

    1990-12-01

    The IAEA Safety Guide 50-SG-S8 devoted to 'Safety Aspects of Foundations of Nuclear Power Plants' indicates that operator of a NPP should establish a program for inspection of safe operation during construction, start-up and service life of the plant for obtaining data needed for estimating the life time of structures and components. At the same time the program should ensure that the safety margins are appropriate. Periodic safety analysis are an important part of the safety inspection program. Periodic safety reports is a method for testing the whole system or a part of the safety system following the precise criteria. Periodic safety analyses are not meant for qualification of the plant components. Separate analyses are devoted to: start-up, qualification of components and materials, and aging. All these analyses are described in this presentation. The last chapter describes the experience obtained for PWR-900 and PWR-1300 units from 1986-1989.

  12. Achieving reasonable conservatism in nuclear safety analyses

    Jamali, Kamiar

    2015-01-01

    In the absence of methods that explicitly account for uncertainties, seeking reasonable conservatism in nuclear safety analyses can quickly lead to extreme conservatism. The rate of divergence to extreme conservatism is often beyond the expert analysts’ intuitive feeling, but can be demonstrated mathematically. Too much conservatism in addressing the safety of nuclear facilities is not beneficial to society. Using certain properties of lognormal distributions for representation of input parameter uncertainties, example calculations for the risk and consequence of a fictitious facility accident scenario are presented. Results show that there are large differences between the calculated 95th percentiles and the extreme bounding values derived from using all input variables at their upper-bound estimates. Showing the relationship of the mean values to the key parameters of the output distributions, the paper concludes that the mean is the ideal candidate for representation of the value of an uncertain parameter. The mean value is proposed as the metric that is consistent with the concept of reasonable conservatism in nuclear safety analysis, because its value increases towards higher percentiles of the underlying positively skewed distribution with increasing levels of uncertainty. Insensitivity of the results to the actual underlying distributions is briefly demonstrated. - Highlights: • Multiple conservative assumptions can quickly diverge into extreme conservatism. • Mathematics and attractive properties provide basis for wide use of lognormal distribution. • Mean values are ideal candidates for representation of parameter uncertainties. • Mean values are proposed as reasonably conservative estimates of parameter uncertainties

  13. Safety analyses for reprocessing and waste processing

    1983-03-01

    Presentation of an incident analysis of process steps of the RP, simplified considerations concerning safety, and safety analyses of the storage and solidification facilities of the RP. A release tree method is developed and tested. An incident analysis of process steps, the evaluation of the SRL-study and safety analyses of the storage and solidification facilities of the RP are performed in particular. (DG) [de

  14. Passive safety injection experiments and analyses (PAHKO)

    Tuunanen, J.

    1998-01-01

    PAHKO project involved experiments on the PACTEL facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines (Pressure Balancing Line, PBL, and Injection Line, IL). The examined PSIS worked efficiently in SBLOCAs although the flow through the PSIS stopped temporarily if the break was very small and the hot water filled the CMT. The experiments demonstrated the importance of the flow distributor in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable to simulate the overall behaviour of the transients. The detailed analyses of the results showed some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the PSIS phenomena. (orig.)

  15. Implementing partnerships in nonreactor facility safety analyses

    Courtney, J.C.; Perry, W.H.; Phipps, R.D.

    1996-01-01

    Faculty and students from LSU have been participating in nuclear safety analyses and radiation protection projects at ANL-W at INEL since 1973. A mutually beneficial relationship has evolved that has resulted in generation of safety-related studies acceptable to Argonne and DOE, NRC, and state regulatory groups. Most of the safety projects have involved the Hot Fuel Examination Facility or the Fuel Conditioning Facility; both are hot cells that receive spent fuel from EBR-II. A table shows some of the major projects at ANL-W that involved LSU students and faculty

  16. Safety analyses for high-temperature reactors

    Mueller, A.

    1978-01-01

    The safety evaluation of HTRs may be based on the three methods presented here: The licensing procedure, the probabilistic risk analysis, and the damage extent analysis. Thereby all safety aspects - from normal operation to the extreme (hypothetical) accidents - of the HTR are covered. The analyses within the licensing procedure of the HTR-1160 have shown that for normal operation and for the design basis accidents the radiation exposures remain clearly below the maximum permissible levels as prescribed by the radiation protection ordinance, so that no real hazard for the population will avise from them. (orig./RW) [de

  17. Demonstration of inherent safety features of HTGRs using the HTTR

    Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

    2004-01-01

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

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

    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)

  19. Reliability and safety analyses under fuzziness

    Onisawa, T.; Kacprzyk, J.

    1995-01-01

    Fuzzy theory, for example possibility theory, is compatible with probability theory. What is shown so far is that probability theory needs not be replaced by fuzzy theory, but rather that the former works much better in applications if it is combined with the latter. In fact, it is said that there are two essential uncertainties in the field of reliability and safety analyses: One is a probabilistic uncertainty which is more relevant for mechanical systems and the natural environment, and the other is fuzziness (imprecision) caused by the existence of human beings in systems. The classical probability theory alone is therefore not sufficient to deal with uncertainties in humanistic system. In such a context this collection of works will put a milestone in the arguments of probability theory and fuzzy theory. This volume covers fault analysis, life time analysis, reliability, quality control, safety analysis and risk analysis. (orig./DG). 106 figs

  20. Thermal hydraulic reactor safety analyses and experiments

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  1. Architecture Level Safety Analyses for Safety-Critical Systems

    K. S. Kushal

    2017-01-01

    Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

  2. Passengers' perception of the safety demonstration on board an aircraft

    Ruenruoy, Ratchada

    The cabin safety demonstration on board an aircraft is one of the methods to provide safety information for passengers before aircraft takeoff. However, passengers' enthusiasm toward safety demonstrations is normally low. Therefore, the study of passengers' perception toward safety briefings on board an aircraft is important in increasing the safety awareness for the travelling public on commercial aircraft. A survey was distributed to measure the perceptions of Middle Tennessee State University (MTSU) faculty and staff, Aerospace students, and international students who have traveled in the last year. It was generally found that watching the cabin safety demonstration before aircraft takeoff was believed to be important for passengers. However, the attention to the safety demonstration remained low because the safety briefings were not good enough in terms of clear communication, particularly in the recorded audio demonstration and the live safety demonstration methods of briefing.

  3. Public education through safety culture demonstration

    Wanitsuksombut, Warapon

    2005-01-01

    The activities relating to nuclear energy have been world widely opposed against, because there have existed scars in the past; atomic bombs and a few accidents in nuclear facilities. It cannot be denied that the most effective education of public is through Medias such as news or documentary on newspaper and television. Once such cases appeared to public, it is difficult to erase the bad pictures from their memory. Since education for public is mainly depending on media, it is recommended putting harder effort on dissemination of information on regulation and regulatory function to public. The regulatory function of each country is the key of safe utilization of nuclear energy. Since prime responsibility of maintenance and operation are rested on the operators. To achieve the goal of safety, regulatory authority's task now is emphasized on encouraging operators of nuclear facilities to implement their safety culture. This will reduce the probability of unwanted events and therefore raising credit of nuclear energy. (author)

  4. 76 FR 24831 - Site-Specific Analyses for Demonstrating Compliance With Subpart C Performance Objectives

    2011-05-03

    ...-level radioactive waste disposal facilities to conduct site-specific analyses to demonstrate compliance... public health and safety, these amendments would enhance the safe disposal of low-level radioactive waste... would be to enhance the safe disposal of low-level radioactive waste. The NRC is also proposing...

  5. Demonstration of safety for geologic disposal

    Taylor, E.C.; Ramspott, L.D.; Sprecher, W.M.

    1994-01-01

    The US Department of Energy (DOE) is developing a nuclear waste management system that will accept high-level radioactive waste, transport it, store it, and ultimately emplace it in a deep geologic repository. The key activity now is determining whether Yucca Mountain, Nevada is suitable as a site for the repository. If so, the crucial technological advance will be the demonstration that disposal of nuclear waste will be safe for thousands of years after closure. This paper assesses the impact of regulatory developments, legal developments, and scientific developments on such a demonstration

  6. Demonstrating the Safety and Reliability of a New System or Spacecraft: Incorporating Analyses and Reviews of the Design and Processing in Determining the Number of Tests to be Conducted

    Vesely, William E.; Colon, Alfredo E.

    2010-01-01

    Design Safety/Reliability is associated with the probability of no failure-causing faults existing in a design. Confidence in the non-existence of failure-causing faults is increased by performing tests with no failure. Reliability-Growth testing requirements are based on initial assurance and fault detection probability. Using binomial tables generally gives too many required tests compared to reliability-growth requirements. Reliability-Growth testing requirements are based on reliability principles and factors and should be used.

  7. Status of the safety concept and safety demonstration for an HLW repository in salt. Summary report

    Bollingerfehr, W.; Buhmann, D.; Filbert, W.; and others

    2013-12-15

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safety assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  8. Progress for the Industry Application External Hazard Analyses Early Demonstration

    Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ryan, Emerald [Idaho State Univ., Pocatello, ID (United States); Bhandari, Bishwo [Idaho State Univ., Pocatello, ID (United States); Sludern, Daniel [Idaho State Univ., Pocatello, ID (United States); Pope, Chad [Idaho State Univ., Pocatello, ID (United States); Sampath, Ram [Centroid PIC, Idaho Falls, ID (United States)

    2015-09-01

    This report describes the current progress and status related to the Industry Application #2 focusing on External Hazards. For this industry application within the Light Water Reactor Sustainability (LWRS) Program Risk-Informed Safety Margin Characterization (RISMC) R&D Pathway, we will create the Risk-Informed Margin Management (RIMM) approach to represent meaningful (i.e., realistic facility representation) event scenarios and consequences by using an advanced 3D facility representation that will evaluate external hazards such as flooding and earthquakes in order to identify, model and analyze the appropriate physics that needs to be included to determine plant vulnerabilities related to external events; manage the communication and interactions between different physics modeling and analysis technologies; and develop the computational infrastructure through tools related to plant representation, scenario depiction, and physics prediction. One of the unique aspects of the RISMC approach is how it couples probabilistic approaches (the scenario) with mechanistic phenomena representation (the physics) through simulation. This simulation-based modeling allows decision makers to focus on a variety of safety, performance, or economic metrics. In this report, we describe the evaluation of various physics toolkits related to flooding representation. Ultimately, we will be coupling the flooding representation with other events such as earthquakes in order to provide coupled physics analysis for scenarios where interactions exist.

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

    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)

  10. Demonstrating safety: Lessons learnt by InSOTEC

    Kallenbach-Herbert, Beate; Brohmann, Bettina

    2014-01-01

    InSOTEC is a three-year collaborative social sciences research project funded under the European Atomic Energy Community's 7. Framework Programme FP7/2007-2011, under grant agreement no. 2699009.1 The project aims to generate a better understanding of the complex interplay between the technical and the social in radioactive waste management (RWM) and, in particular, in the context of the design and implementation of geological disposal. In doing so, InSOTEC wants to move beyond the social and technical division by treating RWM and geological disposal as 'socio-technical' challenges and in following the relationship and describing the context, one can identify the dependency as a socio-technical combination. InSOTEC focuses on situations and issues where the relationship between the technical and social components of geological disposal are still unstable, ambiguous or controversial, and where negotiations are taking place in terms of problem definitions and preferred solutions. Some concrete examples of socio-technical challenges are the question of siting and of introducing the notion of reversibility and retrievability or long-term repository monitoring into the concept of geological disposal. These examples show that the concept of geological disposal develops over time, not only because of evolutions in scientific knowledge, but also as a consequence of debates on how to implement this technology in the light of societal requirements. During the first year of the project, various research activities in the national context of InSOTEC partner countries as well as on the European and international levels contributed to the identification of the main socio-technical challenges in geological disposal. On this basis four topics were selected for in-depth analysis: - reversibility and retrievability; - demonstrating safety; - siting; - technology transfer; The aim of these analyses is to come to a better understanding of the relationships between social and technical

  11. The impact of safety analyses on the design of the Hanford Waste Vitrification Plant

    Koppenaal, T.J.; Yee, A.K.; Reisdorf, J.B.; Hall, B.W.

    1993-04-01

    Accident analyses are being performed to evaluate and document the safety of the Hanford Waste Vitrification Plant (HWVP). The safety of the HWVP is assessed by evaluating worst-case accident scenarios and determining the dose to offsite and onsite receptors. Air dispersion modeling is done with the GENII computer code. Three accidents are summarized in this paper, and their effects on the safety and the design of the HWVP are demonstrated

  12. Technical safety appraisal of the West Valley Demonstration Project

    1989-09-01

    This report presents the results of one in a series of Technical Safety Appraisals (TSAs) being conducted of DOE nuclear operations by the Assistant Secretary for Environment, Safety, and Health Office of Safety Appraisals TSAs are one of the ititiatives announced by the Secretary of Energy on September 18, 1985, to enhance the DOE environment, safety and health program. This report presents the results of a TSA of the West Valley Demonstration Project (WVDP). The appraisal was conducted by a team of exerts assembled by the DOE Office of Safety Appraisal and was conducted during onsite visits of June 26-30 and July 10-21, 1989. West Valley, about 30 miles south of Buffalo, New York is the location of the only commercial nuclear fuel reprocessing facility operated in the United States. Nuclear Fuels Services, Inc. (NFS) operated the plant from 1966 to 1972 and processed about 640 metric tons of spent reactor fuel. The reprocessing operation generated about 560,000 gallons of high-level radioactive waste, which was transferred into underground tanks for storage. In 1972 NFS closed the plant and subsequently decided not to reopen it

  13. Quality assurance requirements for the computer software and safety analyses

    Husarecek, J.

    1992-01-01

    The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref

  14. Integrating scientific results for a post-closure safety demonstration

    Taylor, E.C.; Ramspott, L.D.; Sinnock, S.; Sprecher, W.M.

    1994-01-01

    The U.S. Department of Energy (DOE) is developing a nuclear waste management system that will accept high-level radioactive waste, transport it, store it, and ultimately emplace it in a deep geologic repository. The key activity now is determining whether Yucca Mountain, Nevada is suitable as a site for the repository. If so, the crucial technological advance will be the demonstration that disposal of nuclear waste will be safe for thousands of years after closure. Recent regulatory, legal, and scientific developments imply that the safety demonstration must be simple. The scientific developments taken together support a simple set of hypotheses that constitute a post-closure safety argument for a repository at Yucca Mountain. If the understanding of Yucca Mountain hydrology presented in the Site Characterization Plan proves correct, then these hypotheses might be confirmed by combining results of Surface-Based Testing with early testing results in the Exploratory Studies Facility

  15. Radiation safety at the West Valley Demonstration Project

    Hoffman, R.L.

    1997-01-01

    This is a report on the Radiation Safety Program at the West Valley Demonstration Project (WVDP). This Program covers a number of activities that support high-level waste solidification, stabilization of facilities, and decontamination and decommissioning activities at the Project. The conduct of the Program provides confidence that all occupational radiation exposures received during operational tasks at the Project are within limits, standards, and program requirements, and are as low as reasonably achievable

  16. SCALE Graphical Developments for Improved Criticality Safety Analyses

    Barnett, D.L.; Bowman, S.M.; Horwedel, J.E.; Petrie, L.M.

    1999-01-01

    New computer graphic developments at Oak Ridge National Ridge National Laboratory (ORNL) are being used to provide visualization of criticality safety models and calculational results as well as tools for criticality safety analysis input preparation. The purpose of this paper is to present the status of current development efforts to continue to enhance the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system. Applications for criticality safety analysis in the areas of 3-D model visualization, input preparation and execution via a graphical user interface (GUI), and two-dimensional (2-D) plotting of results are discussed

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

    Delage, M; Giroux, C

    1990-12-01

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

  18. Site-Specific Analyses for Demonstrating Compliance with 10 CFR 61 Performance Objectives - 12179

    Grossman, C.J.; Esh, D.W.; Yadav, P.; Carrera, A.G. [U.S. Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

    2012-07-01

    The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations at 10 CFR Part 61 to require low-level radioactive waste disposal facilities to conduct site-specific analyses to demonstrate compliance with the performance objectives in Subpart C. The amendments would require licensees to conduct site-specific analyses for protection of the public and inadvertent intruders as well as analyses for long-lived waste. The amendments would ensure protection of public health and safety, while providing flexibility to demonstrate compliance with the performance objectives, for current and potential future waste streams. NRC staff intends to submit proposed rule language and associated regulatory basis to the Commission for its approval in early 2012. The NRC staff also intends to develop associated guidance to accompany any proposed amendments. The guidance is intended to supplement existing low-level radioactive waste guidance on issues pertinent to conducting site-specific analyses to demonstrate compliance with the performance objectives. The guidance will facilitate implementation of the proposed amendments by licensees and assist competent regulatory authorities in reviewing the site-specific analyses. Specifically, the guidance provides staff recommendations on general considerations for the site-specific analyses, modeling issues for assessments to demonstrate compliance with the performance objectives including the performance assessment, intruder assessment, stability assessment, and analyses for long-lived waste. This paper describes the technical basis for changes to the rule language and the proposed guidance associated with implementation of the rule language. The NRC staff, per Commission direction, intends to propose amendments to 10 CFR Part 61 to require licensees to conduct site-specific analyses to demonstrate compliance with performance objectives for the protection of public health and the environment. The amendments would require a

  19. Reactor safety research - visible demonstrations and credible computations

    Loewenstein, W B; Divakaruni, S M

    1985-11-01

    EPRI has been conducting nuclear safety research for a number of years with the primary goal of assuring the safety and reliability of the nuclear plants. The visibility is emphasized by sponsoring or participating in large scale test demonstrations to credibly support the complex computations that are the basis for quantification of safety margins. Recognizing the success of the airline industry in receiving favorable public perception, the authors compare the design and operation practices of the airline industry with those of the nuclear industry practices to identify the elements contributing to public concerns and unfavorable perceptions. In this paper, authors emphasize the importance of proper communications of research results to the public in a manner that non-specialists understand. Further, EPRI supported research and results in the areas of source term, seismic and structural engineering research, analysis using probabilistic risk assessment (PRA), quantification of safety margins, digital technology development and implementation, and plant transient and performance evaluations are discussed in the paper. (orig./HP).

  20. Reactor safety research - visible demonstrations and credible computations

    Loewenstein, W.B.; Divakaruni, S.M.

    1985-01-01

    EPRI has been conducting nuclear safety research for a number of years with the primary goal of assuring the safety and reliability of the nuclear plants. The visibility is emphasized by sponsoring or participating in large scale test demonstrations to credibly support the complex computations that are the basis for quantification of safety margins. Recognizing the success of the airline industry in receiving favorable public perception, the authors compare the design and operation practices of the airline industry with those of the nuclear industry practices to identify the elements contributing to public concerns and unfavorable perceptions. In this paper, authors emphasize the importance of proper communications of research results to the public in a manner that non-specialists understand. Further, EPRI supported research and results in the areas of source term, seismic and structural engineering research, analysis using probabilistic risk assessment (PRA), quantification of safety margins, digital technology development and implementation, and plant transient and performance evaluations are discussed in the paper. (orig./HP)

  1. Safety demonstration test on solvent fire in fuel reprocessing plant

    Nishio, Gunji; Hashimoto, Kazuichiro

    1989-03-01

    This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

  2. The role of CFD computer analyses in hydrogen safety management

    Komen, E.M.J; Visser, D.C; Roelofs, F.; Te Lintelo, J.G.T

    2014-01-01

    The risks of hydrogen release and combustion during a severe accident in a light water reactor have attracted considerable attention after the Fukushima accident in Japan. Reliable computer analyses are needed for the optimal design of hydrogen mitigation systems, like e.g. passive autocatalytic recombiners (PARs), and for the assessment of the associated residual risk of hydrogen combustion. Traditionally, so-called Lumped Parameter (LP) computer codes are being used for these purposes. In the last decade, significant progress has been made in the development, validation, and application of more detailed, three-dimensional Computational Fluid Dynamics (CFD) simulations for hydrogen safety analyses. The objective of the current paper is to address the following questions: - When are CFD computer analyses needed complementary to the traditional LP code analyses for hydrogen safety management? - What is the validation status of the CFD computer code for hydrogen distribution, mitigation, and combustion analyses? - Can CFD computer analyses nowadays be executed in practical and reliable way for full scale containments? The validation status and reliability of CFD code simulations will be illustrated by validation analyses performed for experiments executed in the PANDA, THAI, and ENACCEF facilities. (authors)

  3. Demonstration tests for low level radioactive waste packaging safety

    Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

    1993-01-01

    The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

  4. Experience on the demonstration of safety for older reactors

    Facer, R.

    2001-01-01

    The UK's oldest reactors are still operating. Built during the 1950's and commissioned between 1956 and 1960, eight reactors continue to provide electricity and process steam. It is still economically justified to keep them running. In addition to the economic considerations it is also necessary to justify that they can still continue to operate safely. This paper provides a brief review of how the Operator of these stations has justified the safety of operation to date and how they expect to continue to justify their operation for several more years. It is appropriate to consider why the Operator wishes to keep the plant operating. Among the most important reasons are that: The plant is built and paid for, Running costs are relatively low process steam is available for the adjacent sites It is a commercially viable electricity producer It is a reliable electricity source The operators have developed programmes for safety review of the plant and introduced a Continuing Operation Programme which had two main requirements which were, the demonstration of continuing acceptable safety the ensurance of commercial viability. (author)

  5. Demonstrating a correlation between the maturity of road safety practices and road safety incidents.

    Amador, Luis; Willis, Christopher Joseph

    2014-01-01

    The objective of this study is to demonstrate a correlation between the maturity of a country's road safety practices and road safety incidents. Firstly, data on a number of road injuries and fatalities for 129 countries were extracted from the United Nations Global Status on Road Safety database. These data were subdivided according to road safety incident and accident causation factors and normalized based on vehicular fleet (per 1000 vehicles) and road network (per meter of paved road). Secondly, a road safety maturity model was developed based on an adaptation of the concept of process maturity modeling. The maturity of countries with respect to 10 road safety practices was determined through the identification of indicators recorded in the United Nations Global Status of Road Safety Database. Plots of normalized road safety performance of the 129 countries against their maturity scores for each road safety practice as well as an aggregation of the road safety practices were developed. An analysis of variance was done to determine the extent of the correlation between the road safety maturity of the countries and their performance. In addition, a full Bayesian analysis was done to confirm the correlation of each of the road safety practices with injuries and fatalities. Regression analysis for fatalities, injuries, and combined accidents identified maturity with respect to road safety practices associated with speed limits and use of alternative modes as being the most significant predictors of traffic fatalities. A full Bayesian regression confirms that there is a correlation between the maturity of road safety practices and road safety incidents. Road safety practices associated with enforcement of speed limits and promotion of alternative modes are the most significant road safety practices toward which mature countries have concentrated their efforts, resulting in a lower frequency of fatalities, injury rates, and property damage accidents. The authors

  6. Demonstration of safety of decommissioning of facilities using radioactive material

    Batandjieva, Borislava; O'Donnell, Patricio

    2008-01-01

    Full text:The development of nuclear industry worldwide in the recent years has particular impact on the approach of operators, regulators and interested parties to the implementation of the final phases (decommissioning) of all facilities that use radioactive material (from nuclear power plants, fuel fabrication facilities, research reactors to small research or medical laboratories). Decommissioning is becoming an increasingly important activity for two main reasons - termination of the practice in a safe manner with the view to use the facility or the site for other purposes, or termination of the practice and reuse the facility or site for new built nuclear facilities. The latter is of special relevance to multi-facility sites where for example new nuclear power plants and envisaged. However, limited countries have the adequate legal and regulatory framework, and experience necessary for decommissioning. In order to respond to this challenge of the nuclear industry and assist Member States in the adequate planning, conduct and termination of decommissioning of wide range of facilities, over the last decade the IAEA has implemented and initiated several projects in this field. One of the main focuses of this assistance to operators, regulators and specialists involved in decommissioning is the evaluation and demonstration of safety of decommissioning. This importance of these Agency activities was also highlighted in the International Action Plan on Decommissioning, during the second Joint Convention meeting in 2006 and the International Conference on Lessons Learned from Decommissioning in Athens in 2006. The IAEA has been providing technical support to its Member States in this field through several mechanisms: (1) the establishment of a framework of safety standards on decommissioning and development of a supporting technical documents; (2) the establishment of an international peer review mechanism for decommissioning; (3) the technical cooperation projects

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

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

    2004-01-01

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

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

    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

  9. RETRAN safety analyses of the nuclear-powered ship Mutsu

    Yoshinori, N.; Ishida, T.; Tanaka, Y.; Yoshiaki, F.

    1983-01-01

    A number of operational transient analyses of the nuclear-powered ship Mutsu have been performed in response to Japanese nuclear safety regulatory concerns. The RETRAN and COBRA-IV computer codes were used to provide a quantitative basis for the safety evaluation of the plant. This evaluation includes a complete loss of load without reactor scram, an excessive load increase incident, and an accidental depressurization of the primary system. The minimum departure from nucleate boiling ratio remained in excess of 1.53 for these three transients. Hence, the integrity of the core was shown to be maintained during these transients. Comparing the transient behaviors with those of land-based pressurized water reactors, the characteristic features of the Mutsu reactor were presented and the safety of the plant under the operational transient conditions was confirmed

  10. Nuclear power plants: Results of recent safety analyses

    Steinmetz, E.

    1987-01-01

    The contributions deal with the problems posed by low radiation doses, with the information currently available from analyses of the Chernobyl reactor accident, and with risk assessments in connection with nuclear power plant accidents. Other points of interest include latest results on fission product release from reactor core or reactor building, advanced atmospheric dispersion models for incident and accident analyses, reliability studies on safety systems, and assessment of fire hazard in nuclear installations. The various contributions are found as separate entries in the database. (DG) [de

  11. Use of probabilistic safety analyses in severe accident management

    Neogy, P.; Lehner, J.

    1991-01-01

    An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs

  12. Safety analyses of the electrical systems on VVER NPP

    Andel, J.

    2004-01-01

    Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

  13. Safety assessment for the 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generation

    Suk, Ho Chun; Cho, M. S.; Jun, J. S. and others

    2001-06-01

    This document is a report on the safety assessment for the 24 CANFLEX-NU(CANDU Flexible fuelling - Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station. The CANFLEX fuel bundle as a CANDU advanced fuel has been jointly developed by KAERI/AECL. This document describes the rationale for the demonstration irradiation and comments on the Korean government licensing issues such as the status of the CANFLEX fuel irradiations at NRU research reactor in AECL, status and plan of the CANFLEX fuel irradiations at a CANDU-6 power reactor, status of the water CHF(Critical Heat Flux) test at Stern Laboratories and the CHF correlation. This documents presents an assessment the consequences of postulated accidents with all safety system available during demonstration irradiation of 24 CANFLEX-NU fuel bundles at Wolsong-1 Generating Station. The assessment is made by two kinds of approaches. One approach is based on the document of the safety assessment for the 24 CANFLEX-NU fuel bundle demonstration irradiation at Point Lepreau Generating Station. The other approach is taken from the safety analyses using the analysis methods and assumptions used in the final safety reports on the 600 MWe CANDU-PHWR Wolsung-2, 3, and 4 Nuclear Power Plants for the Korea Electric Power Cooperation. The analyses are not comprehensive reviews of the postulated accidents, but examination of the expected difference in accident consequences because of the presence of 24 CANFLEX fuel bundles in two channels. The approach is to compare the difference to the safety margin for 37-element bundle cases.

  14. Criticality safety analyses in SKODA JS a.s

    Mikolas, P.; Svarny, J.

    1999-01-01

    This paper describes criticality safety analyses of spent fuel systems for storage and transport of spent fuel performed in SKODA JS s.r.o.. Analyses were performed for different systems both at NPP site including originally designed spent fuel pool with a large pitch between assemblies without any special absorbing material, high density spent fuel pool with an additional absorption by boron steel, depository rack for fresh fuel assemblies with a very large pitch between fuel assemblies, a container for transport of fresh fuel into the reactor pool and a cask for transport and storage of spent fuel and container for final storage depository. required subcriticality has been proven taking into account all possible unfavourable conditions, uncertainties etc. In two cases, burnup credit methodology is expected to be used. (Authors)

  15. Method of accounting for code safety valve setpoint drift in safety analyses

    Rousseau, K.R.; Bergeron, P.A.

    1989-01-01

    In performing the safety analyses for transients that result in a challenge to the reactor coolant system (RCS) pressure boundary, the general acceptance criterion is that the peak RCS pressure not exceed the American Society of Mechanical Engineers limit of 110% of the design pressure. Without crediting non-safety-grade pressure mitigating systems, protection from this limit is mainly provided by the primary and secondary code safety valves. In theory, the combination of relief capacity and setpoints for these valves is designed to provide this protection. Generally, banks of valves are set at varying setpoints staggered by 15- to 20-psid increments to minimize the number of valves that would open by an overpressure challenge. In practice, however, when these valves are removed and tested (typically during a refueling outage), setpoints are sometimes found to have drifted by >50 psid. This drift should be accounted for during the performance of the safety analysis. This paper describes analyses performed by Yankee Atomic Electric Company (YAEC) to account for setpoint drift in safety valves from testing. The results of these analyses are used to define safety valve operability or acceptance criteria

  16. Safety design analyses of Korea Advanced Liquid Metal Reactor

    Suk, S.D.; Park, C.K.

    2000-01-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

  17. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    Aydogdu, K.; Boss, C. R.

    2006-01-01

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  18. Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

    Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

    2011-01-01

    This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

  19. Requirements on the provisional safety analyses and technical comparison of safety measures

    2010-04-01

    The concept of a Geological Underground Repository (SGT) was adopted by the Swiss Federal Council on April 2 nd , 2008. It fixes the goals and the safety technical criteria as well as the procedures for the choice of the site for an underground repository. Those responsible for waste management evaluate possible site regions according to the present status of geological knowledge and based on the safety criteria defined in SGT as well as on technical feasibility. In a first step, they propose geological repository sites for high level (HAA) and for low and intermediate level (SMA) radioactive wastes and justify their choice in a report delivered to the Swiss Federal Office of Energy. The Swiss Federal Council reviews the choices presented and, in the case of positive evaluation, approves them and considers them as an initial orientation. In a second step, based on the possible sites according to step 1, the waste management institution responsible has to reduce the repositories chosen for HAA and SMA by taking into account safety aspects, technical feasibility as well as space planning and socio-economical aspects. In making this choice, safety aspects have the highest priority. The criteria used for the evaluation in the first step have to be defined using provisional quantitative safety analyses. On the basis of the whole appraisal, including space planning and socio-economical aspects, those responsible for waste management propose at least two repository sites for HAA- and SMA-waste. Their selection is then reviewed by the authorities and, in the case of a positive assesment, the selection is taken as an intermediate result. The remaining sites are further studied to examine site choice and the delivery of a request for a design license. If necessary, the requested geological knowledge has to be confirmed by new investigations. Based on the results of the choosing process and a positive evaluation by the safety authorities, the Swiss Federal Council has to

  20. Methodology development for statistical evaluation of reactor safety analyses

    Mazumdar, M.; Marshall, J.A.; Chay, S.C.; Gay, R.

    1976-07-01

    In February 1975, Westinghouse Electric Corporation, under contract to Electric Power Research Institute, started a one-year program to develop methodology for statistical evaluation of nuclear-safety-related engineering analyses. The objectives of the program were to develop an understanding of the relative efficiencies of various computational methods which can be used to compute probability distributions of output variables due to input parameter uncertainties in analyses of design basis events for nuclear reactors and to develop methods for obtaining reasonably accurate estimates of these probability distributions at an economically feasible level. A series of tasks was set up to accomplish these objectives. Two of the tasks were to investigate the relative efficiencies and accuracies of various Monte Carlo and analytical techniques for obtaining such estimates for a simple thermal-hydraulic problem whose output variable of interest is given in a closed-form relationship of the input variables and to repeat the above study on a thermal-hydraulic problem in which the relationship between the predicted variable and the inputs is described by a short-running computer program. The purpose of the report presented is to document the results of the investigations completed under these tasks, giving the rationale for choices of techniques and problems, and to present interim conclusions

  1. Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

    Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-10-01

    Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.

  2. Demonstration of the LHC Safety Training Tunnel Mock-Up

    Brice, Maximilien

    2014-01-01

    Members of CERN's management visit the LHC tunnel mock-up at the Safety Training Centre on the Prévessin site. The facility is used to train personnel in emergency responses including the use of masks and safe evacuation.

  3. The HTR safety concept demonstrated by selected examples

    Sommer, H.; Stoelzl, D.

    1981-01-01

    The licensing experience gained in the Federal Republic of Germany is based on the licensing procedures for the THTR-300 and the HTR-1160. In the course of the licensing procedures for these reactors a safety concept for an HTR has been developed. This experience constitutes the basis for the design of future HTR's. (author)

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

    Becker, D.L.

    1994-11-01

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

  5. Radioactive waste management in France: safety demonstration fundamentals.

    Ouzounian, G; Voinis, S; Boissier, F

    2012-01-01

    The main challenge in development of the safety case for deep geological disposal is associated with the long periods of time over which high- and intermediate-level long-lived wastes remain hazardous. A wide range of events and processes may occur over hundreds of thousands of years. These events and processes are characterised by specific timescales. For example, the timescale for heat generation is much shorter than any geological timescale. Therefore, to reach a high level of reliability in the safety case, it is essential to have a thorough understanding of the sequence of events and processes likely to occur over the lifetime of the repository. It then becomes possible to assess the capability of the repository to fulfil its safety functions. However, due to the long periods of time and the complexity of the events and processes likely to occur, uncertainties related to all processes, data, and models need to be understood and addressed. Assessment is required over the lifetime of the radionuclides contained in the radioactive waste. Copyright © 2012. Published by Elsevier Ltd.

  6. Industrial Fuel Gas Demonstration Plant Program. Task III, Demonstration plant safety, industrial hygiene, and major disaster plan (Deliverable No. 35)

    None

    1980-03-01

    This Health and Safety Plan has been adopted by the IFG Demonstration Plant managed by Memphis Light, Gas and Water at Memphis, Tennessee. The plan encompasses the following areas of concern: Safety Plan Administration, Industrial Health, Industrial Safety, First Aid, Fire Protection (including fire prevention and control), and Control of Safety Related Losses. The primary objective of this plan is to achieve adequate control of all potentially hazardous activities to assure the health and safety of all employees and eliminate lost work time to both the employees and the company. The second objective is to achieve compliance with all Federal, state and local laws, regulations and codes. Some thirty specific safe practice instruction items are included.

  7. Demonstration of the reliability of the safety pumps

    Durand, J.M.

    1989-01-01

    POMPES GUINARD is supplying about 60% of the Nuclear pumps for the French Program. To become the specialist of Safety Related Pumps POMPES GUINARD made a lot of efforts and investments to acquire knowledge and experience. This was possible mainly with test on special loops as it is the only way for a pump manufacturer to progress by controlling hydraulics, components, bearings, mechanical seals, inducer, mechanical and hydraulic behaviour of the units in process of time. We will describe hereafter some of the typical tests which were performed during the last fifteen years

  8. Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

    Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (238Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (237Np) dioxide (NpO2)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

  9. Demonstration of passive safety features in EBR-II

    Planchon, H.P. Jr.; Golden, G.H.; Sackett, J.I.

    1987-01-01

    Two tests of great importance to the design of future commercial nuclear power plants were carried out in the Experimental Breeder Reactor-II on April 3, 1986. These tests, (viewed by about 60 visitors, including 13 foreign LMR specialists) were a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. In these tests, inherent feedback shut the reactor down without damage to the fuel or other reactor components. This resulted primarily from advantageous characteristics of the metal driver fuel used in EBR-II. Work is currently underway at EBR-II to develop a control strategy that promotes inherent safety characteristics, including survivability of transient overpower accidents. In parallel, work is underway at EBR-II on the development of state-of-the-art plant diagnostic techniques

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

    Lipar, M.

    1997-01-01

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

  11. Safety analysis report for packaging (onsite) transuranic performance demonstration program sample packaging

    Mccoy, J.C.

    1997-01-01

    The Transuranic Performance Demonstration Program (TPDP) sample packaging is used to transport highway route controlled quantities of weapons grade (WG) plutonium samples from the Plutonium Finishing Plant (PFP) to the Waste Receiving and Processing (WRAP) facility and back. The purpose of these shipments is to test the nondestructive assay equipment in the WRAP facility as part of the Nondestructive Waste Assay PDP. The PDP is part of the U. S. Department of Energy (DOE) National TRU Program managed by the U. S. Department of Energy, Carlsbad Area Office, Carlsbad, New Mexico. Details of this program are found in CAO-94-1045, Performance Demonstration Program Plan for Nondestructive Assay for the TRU Waste Characterization Program (CAO 1994); INEL-96/0129, Design of Benign Matrix Drums for the Non-Destructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996a); and INEL-96/0245, Design of Phase 1 Radioactive Working Reference Materials for the Nondestructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996b). Other program documentation is maintained by the national TRU program and each DOE site participating in the program. This safety analysis report for packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the TRU PDP sample packaging meets the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for an onsite Transportation Hazard Indicator (THI) 2 packaging. This SARP, however, does not include evaluation of any operations within the PFP or WRAP facilities, including handling, maintenance, storage, or operating requirements, except as they apply directly to transportation between the gate of PFP and the gate of the WRAP facility. All other activities are subject to the requirements of the facility safety analysis reports (FSAR) of the PFP or WRAP facility and requirements of the PDP

  12. Cost/benefit analyses of reactor safety systems

    1988-01-01

    The study presents a methodology for quantitative assessment of the benefit yielded by the various engineered safety systems of a nuclear reactor containment from the standpoint of their capacity to protect the environment compared to their construction costs. The benefit is derived from an estimate of the possible damage from which the environment is protected, taking account of the probabilities of occurrence of malfunctions and accidents. For demonstration purposes, the methodology was applied to a 1 300-MWe PWR nuclear power station. The accident sequence considered was that of a major loss-of-coolant accident as investigated in detail in the German risk study. After determination of the benefits and cost/benefit ratio for the power plant and the containment systems as designed, the performance characteristics of three subsystems, the leakoff system, annulus exhaust air handling system and spray system, were varied. For this purpose, the parameters which describe these systems in the activity release programme were altered. The costs were simultaneously altered in order to take account of the performance divergences. By varying the performance of the individual sub-systems an optimization in design of these systems can be arrived at

  13. Analyses to demonstrate the thermal performance of the CASTOR KN12

    Diersch, R.; Weiss, M.; Tso, C.F.; Powell, D.; Choy, B.I.; Lee, H.Y.

    2004-01-01

    The CASTOR registered KN-12 is a new cask design of GNB for dry and wet transportation of up to 12 PWR spent nuclear fuel assemblies in Korea. It complies with the requirements of 10 CFR 71 [1] and IAEA ST-1 [2] for TYPE B(U)F packages. It received its transport license from the Korean Competent Authority KINS in July 2002 and is now in use in South Korea. Demonstration of the cask's compliance with the regulatory requirements in the area of thermal performance has been carried out by a combination of testing carried out by Korea Atomic Energy Research Institute and analyses carried out by Arup. This paper describes the analyses to demonstrate the thermal performance of the cask and compliance with regulatory requirements under normal and hypothetical accident conditions of transport. Other aspects of the design of the CASTOR registered KN12 are presented in other papers at this conference

  14. Simulator platform for fast reactor operation and safety technology demonstration

    Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-01-01

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  15. Simulator platform for fast reactor operation and safety technology demonstration

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  16. Quantified safety objectives in high technology: Meaning and demonstration

    Vinck, W.F.; Gilby, E.; Chicken, J.

    1986-01-01

    An overview and trends-analysis is given of the types of quantified criteria and objectives which are presently applied or envisaged and discussed in Europe in the nuclear application, more specifically Nuclear Power Plants (NPPs), and in non-nuclear applications, more specifically in the chemical and petrochemical process industry. Some comparative deductions are made. Attention is paid to the similarities or discrepancies between such criteria and objectives and to problems associated with the demonstration that they are implemented. The role of cost-effectiveness of Risk deduction is briefly discussed and mention made of a search made into combining the technical, economic and socio-political factors playing a role in Risk acceptance

  17. Best Estimate plus Uncertainty (BEPU) Analyses in the IAEA Safety Standards

    Dusic, Milorad; )

    2013-01-01

    The Safety Standards Series establishes an essential basis for safety and represents the broadest international consensus. Safety Standards Series publications are categorized into: Safety Fundamental (Present the overall objectives, concepts and principles of protection and safety, they are the policy documents of the safety standards), Safety Requirements (Establish requirements that must be met to ensure the protection and safety of people and the environment, both now and in the future), and Safety Guides (Provide guidance, in the form of more detailed actions, conditions or procedures that can be used to comply with the Requirements). The incorporation of more detailed requirements, in accordance with national practice, may still be necessary. There should be only one set of international safety standards. Each safety standard will be reviewed by the relevant committee or by the commission every five years. Best Estimate plus Uncertainty (BEPU) Analyses are approached in the following IAEA Safety Standards: - Safety Requirements SSR 2/1 - Safety of NPPs, Design (Revision of NS-R-1); - General Safety Requirement GSR Part 4: Safety Assessment for Facilities and Activities; - Safety Guide SSG-2 Deterministic Safety Analysis for Nuclear Power Plants. NUSSC suggested that new safety guides should be accompanied by documents like TECDOCs or Safety Reports describing in detail their recommendations where appropriate. Special review is currently underway to identify needs for revision in the light of the Fukushima accident. Revision will concern, first, the Safety Requirements, and then, the Selected Safety Guides

  18. Use of the deterministic safety analyses in support to the NPP Krsko modification

    Feretic, D.; Cavlina, N.; Debrecin, N.; Grgic, D.; Bajs, T.; Spalj, S.

    2004-01-01

    The ultimate goal of the safety analysis is to verify that Nuclear Power Plant (NPP) meets safety and operational requirements. To this aim it is necessary to demonstrate that plant safety has not been deteriorated in the case of the modifications to the plant Systems, Structures and Components (SSC) or changes to the plant procedures. In addition, safety analyses are needed in the case of reassessment of an existing plant. The reasons for reassessment may be different, e.g. due to the changes in the methodology and assumptions used in the original design, if the original design basis or acceptance criteria may no longer be adequate, if the safety analysis tools used may have been superseded by more sophisticated methods or if the original design basis may no longer be met. The operation of the NPP Krsko has experienced numerous changes from the original design for the majority of the reasons that have been mentioned before. On the other side, the application of the large best-estimate thermalhydraulic codes has evolved to the wide spread support in the operation of the NPP: compliance with the regulatory goals, support to the PSA studies, analysis of the operational transients, plant modifications studies, equipment qualification, training of the operators, preparation of the operating procedures, etc. This trend has been followed at the Faculty of Electrical Engineering Zagreb (FER) and applied to the on-going needs due to the modifications and changes at NPP Krsko. In this paper, an overview of the deterministic safety analyses performed at FER in the support to the NPP Krsko modifications and changes is presented.(author)

  19. Nuclear safety: operational aspects. 1. Demonstrating the Link Between Safety Culture and Competitiveness

    Chakoff, H. Elliot; Slider, James E.

    2001-01-01

    More than 20 years ago, we demonstrated a methodology for distinguishing the safety cultures of nuclear power plants. Using the content of licensee event reports, the methodology led to the identification of metrics that could be used to partition 12 pilot plants into better and poorer performers. The partitioning was validated by U.S. Nuclear Regulatory Commission (NRC) experts and shown to be statistically significant at the 95% level of confidence. We wanted to know if the passage of time had validated the differences in performance identified by the original methodology. Our follow-up confirmed the validity of the methodology and also revealed an order of magnitude difference in the long-term survival probability of the 12 pilot plants. The lessons learned from these studies could help plant owners improve safety culture and competitiveness in today's Darwinian marketplace. The original study sought to determine if it was possible to distinguish between better- and poorer-performing plants. The study found it was possible and developed a methodology for doing so. Key breakthroughs included the following: 1. recognizing that safety performance is a stochastic process; thus, performance data must be evaluated using appropriate methods; 2. developing a model for interpreting and transforming raw data into a root-cause domain; 3. maintaining a rigorous model design logic and selecting analytical tools and procedures consistent with that logic; 4. determining that the number of low significance events is an unreliable measure of performance; 5. recognizing that it is the relationship between events that is crucial to understanding performance and risk. Metrics were developed using a test population of 12 plants selected and grouped as 'good' or 'poor' performers by NRC's most senior inspectors. The test population included three plants that had significant events in a 2-yr period and nine that had none. The metrics validated differences in performance hypothesized

  20. Development of the evaluation methods in reactor safety analyses and core characteristics

    NONE

    2013-08-15

    In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  1. Safety and licensing analyses for the Fort St. Vrain HTGR

    Ball, S.J.; Conklin, J.C.; Harrington, R.M.; Cleveland, J.C.; Clapp, N.E. Jr.

    1982-01-01

    The Oak Ridge National Laboratory (ORNL) safety analysis program for the HTGR includes development and verification of system response simulation codes, and applications of these codes to specific Fort St. Vrain reactor licensing problems. Licensing studies addressed the oscillation problems and the concerns about large thermal stresses in the core support blocks during a postulated accident

  2. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  3. Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003. (author)

  4. Additional methodology development for statistical evaluation of reactor safety analyses

    Marshall, J.A.; Shore, R.W.; Chay, S.C.; Mazumdar, M.

    1977-03-01

    The project described is motivated by the desire for methods to quantify uncertainties and to identify conservatisms in nuclear power plant safety analysis. The report examines statistical methods useful for assessing the probability distribution of output response from complex nuclear computer codes, considers sensitivity analysis and several other topics, and also sets the path for using the developed methods for realistic assessment of the design basis accident

  5. PA activity by using nuclear power plant safety demonstration and analysis

    Tsuchiya, Mitsuo; Kamimae, Rie

    1999-01-01

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

  6. Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

    Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

  7. Multivariate statistical analyses demonstrate unique host immune responses to single and dual lentiviral infection.

    Sunando Roy

    2009-10-01

    Full Text Available Feline immunodeficiency virus (FIV and human immunodeficiency virus (HIV are recently identified lentiviruses that cause progressive immune decline and ultimately death in infected cats and humans. It is of great interest to understand how to prevent immune system collapse caused by these lentiviruses. We recently described that disease caused by a virulent FIV strain in cats can be attenuated if animals are first infected with a feline immunodeficiency virus derived from a wild cougar. The detailed temporal tracking of cat immunological parameters in response to two viral infections resulted in high-dimensional datasets containing variables that exhibit strong co-variation. Initial analyses of these complex data using univariate statistical techniques did not account for interactions among immunological response variables and therefore potentially obscured significant effects between infection state and immunological parameters.Here, we apply a suite of multivariate statistical tools, including Principal Component Analysis, MANOVA and Linear Discriminant Analysis, to temporal immunological data resulting from FIV superinfection in domestic cats. We investigated the co-variation among immunological responses, the differences in immune parameters among four groups of five cats each (uninfected, single and dual infected animals, and the "immune profiles" that discriminate among them over the first four weeks following superinfection. Dual infected cats mount an immune response by 24 days post superinfection that is characterized by elevated levels of CD8 and CD25 cells and increased expression of IL4 and IFNgamma, and FAS. This profile discriminates dual infected cats from cats infected with FIV alone, which show high IL-10 and lower numbers of CD8 and CD25 cells.Multivariate statistical analyses demonstrate both the dynamic nature of the immune response to FIV single and dual infection and the development of a unique immunological profile in dual

  8. Multi-person and multi-attribute design evaluations using evidential reasoning based on subjective safety and cost analyses

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

    1996-01-01

    This paper presents an approach for ranking proposed design options based on subjective safety and cost analyses. Hierarchical system safety analysis is carried out using fuzzy sets and evidential reasoning. This involves safety modelling by fuzzy sets at the bottom level of a hierarchy and safety synthesis by evidential reasoning at higher levels. Fuzzy sets are also used to model the cost incurred for each design option. An evidential reasoning approach is then employed to synthesise the estimates of safety and cost, which are made by multiple designers. The developed approach is capable of dealing with problems of multiple designers, multiple attributes and multiple design options to select the best design. Finally, a practical engineering example is presented to demonstrate the proposed multi-person and multi-attribute design selection approach

  9. Safety analyses of the ARIES tokamak reactor designs

    Herring, J.S.; McCarthy, K.A.; Dolan, T.J.

    1994-01-01

    The ARIES design has sought to maximize environmental and safety advantages of fusion through careful selection of materials and design. The ARIES-I tokamak reactor design consists of an SiC composite structure for the first wall and blanket, cooled by 10MPa helium. The breeder is Li 2 ZrO 3 . The divertor consists of SiC composite tubes coated with 2mm tungsten. Loss-of-cooling accident (LOCA) calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. The ARIES-II design includes liquid lithium and vanadium, both of which have low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe with a total 1km early dose of about 88rem (0.88Sv). ARIES-III was an extensive examination of the viability of a D- 3 He fueled tokamak power reactor. Because neutrons are produced only through side reactions (D+D→ 3 He+n, and D+D→T+p followed by D+T→ 4 He+n), the reactor has a reduced activation of the first wall and shield, low afterheat and class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. We modeled a LOCA in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, below 600 C, release fractions are small. We analyzed the disposition of the 20g per day of tritium that is produced by D-D reactions and removed by vacuum pumps. The ARIES-IV coolant is helium and the breeder is lithium oxide. The structure is silicon carbide. Since the neutron multiplier, beryllium metal, is combustible, releasing about 60MJkg -1 , beryllium is the chief source of chemical energy. Less than 10% of the 24 Na inventory is likely to diffuse out of the SiC during a fire in which the beryllium is consumed. Therefore, the offsite dose would be less than 200rem. ((orig.))

  10. Social learning of fear and safety is determined by the demonstrator's racial group.

    Golkar, Armita; Castro, Vasco; Olsson, Andreas

    2015-01-01

    Social learning offers an efficient route through which humans and other animals learn about potential dangers in the environment. Such learning inherently relies on the transmission of social information and should imply selectivity in what to learn from whom. Here, we conducted two observational learning experiments to assess how humans learn about danger and safety from members ('demonstrators') of an other social group than their own. We show that both fear and safety learning from a racial in-group demonstrator was more potent than learning from a racial out-group demonstrator. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

  11. SOCR Analyses: Implementation and Demonstration of a New Graphical Statistics Educational Toolkit

    Annie Chu

    2009-04-01

    Full Text Available The web-based, Java-written SOCR (Statistical Online Computational Resource toolshave been utilized in many undergraduate and graduate level statistics courses for sevenyears now (Dinov 2006; Dinov et al. 2008b. It has been proven that these resourcescan successfully improve students' learning (Dinov et al. 2008b. Being rst publishedonline in 2005, SOCR Analyses is a somewhat new component and it concentrate on datamodeling for both parametric and non-parametric data analyses with graphical modeldiagnostics. One of the main purposes of SOCR Analyses is to facilitate statistical learn-ing for high school and undergraduate students. As we have already implemented SOCRDistributions and Experiments, SOCR Analyses and Charts fulll the rest of a standardstatistics curricula. Currently, there are four core components of SOCR Analyses. Linearmodels included in SOCR Analyses are simple linear regression, multiple linear regression,one-way and two-way ANOVA. Tests for sample comparisons include t-test in the para-metric category. Some examples of SOCR Analyses' in the non-parametric category areWilcoxon rank sum test, Kruskal-Wallis test, Friedman's test, Kolmogorov-Smirno testand Fligner-Killeen test. Hypothesis testing models include contingency table, Friedman'stest and Fisher's exact test. The last component of Analyses is a utility for computingsample sizes for normal distribution. In this article, we present the design framework,computational implementation and the utilization of SOCR Analyses.

  12. Demonstration of criticality safety for the modified TN-REG and TN-BRP transport/storage casks

    Parks, C.V.; Fox, P.B.

    1989-01-01

    An inability to model the structural performance of borated steel baskets under accident conditions forced the specially designed TN-BRP and TN-REG casks to be modified for half-loaded shipments. This paper discusses the approach used to demonstrate that the half-loaded casks would remain safely subcritical even if no credit were taken for the borated basket. Normal and accident configurations were analyzed with the KENO V.a code. The strategy conceived and the analyses performed to demonstrate an acceptable margin of safety are discussed. 5 refs., 3 figs., 2 tabs

  13. Occupational Safety and Health Program at the West Valley Demonstration Project

    L. M. Calderon

    1999-01-01

    The West Valley Nuclear Services Co. LLC (WVNS) is committed to provide a safe, clean, working environment for employees, and to implement U.S. Department of Energy (DOE) requirements affecting worker safety. The West Valley Demonstration Project (WVDP) Occupational Safety and Health Program is designed to protect the safety, health, and well-being of WVDP employees by identifying, evaluating, and controlling biological, chemical, and physical hazards in the work place. Hazards are controlled within the requirements set forth in the reference section at the end of this report. It is the intent of the WVDP Occupational Safety and Health Program to assure that each employee is provided with a safe and healthy work environment. This report shows the logical path toward ensuring employee safety in planning work at the WVDP. In general, planning work to be performed safely includes: combining requirements from specific programs such as occupational safety, industrial hygiene, radiological control, nuclear safety, fire safety, environmental protection, etc.; including WVDP employees in the safety decision-making processes; pre-planning using safety support re-sources; and integrating the safety processes into the work instructions. Safety management principles help to define the path forward for the WVDP Occupational Safety and Health Program. Roles, responsibilities, and authority of personnel stem from these ideals. WVNS and its subcontractors are guided by the following fundamental safety management principles: ''Protection of the environment, workers, and the public is the highest priority. The safety and well-being of our employees, the public, and the environment must never be compromised in the aggressive pursuit of results and accomplishment of work product. A graded approach to environment, safety, and health in design, construction, operation, maintenance, and deactivation is incorporated to ensure the protection of the workers, the public, and the environment

  14. Integration of safety culture in transient analyses for nuclear power plants

    Stosic, Zoran V.; Stoll, Uwe

    2009-01-01

    In the nuclear field Safety Culture is the arrangement of attitudes and characteristics in individuals and organisations which determines first and foremost that nuclear power plant safety issues receive adequate attention due to their outstanding significance. It differs from general Corporate Culture via its concept of core hazards and the potentially large effects associated with the release of radioactivity. One can talk about positive and negative Safety Cultures. A positive Safety Culture assumes that the whole is more than the sum of the parts. The different parts interact to increase the overall effectiveness. In a negative Safety Culture the opposite is the case, with the action of some individuals restricted by the cynicism of others. Some examples of issues that contribute to a negative safety culture are: non-adherence to the established instructions and procedures, unclear definition of responsibilities, disinterest and inattentiveness, overestimation of own capabilities and arrogance, unclear rules, and mistrust between involved organisations. In addition to differentiation and importance of Safety Culture, necessary commitment levels, safety management framework, the paper discusses integration of Safety Culture in transient analyses of nuclear power plants. In this course the commitment to Safety Culture is defined as: a good Safety Culture depends on the continuous commitment and fulfilment of all involved organizations, persons and processes without any exception. (author)

  15. Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

    Wetzel, N.; Scharfe, A.

    1998-01-01

    The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

  16. Specific experiments carried out in Germany in order to demonstrate the safety of existing structures

    Krutzik, Norbert

    2002-01-01

    Specific experiments are carried out in Germany in order to demonstrate the safety of existing NPPs. HDR research program includes operational loads testing (pressure test, pressure and temperature test, thermal shock, fatigue); extreme loads (earthquake, aircraft crash, external explosion); internal emergency loads (blowdown, hydrogen combustion, fire, thermal shock, water hammer, condensation loads)

  17. Application of geostatistical methods to long-term safety analyses for radioactive waste repositories

    Roehlig, K.J.

    2001-01-01

    Long-term safety analyses are an important part of the design and optimisation process as well as of the licensing procedure for final repositories for radioactive waste in deep geological formations. For selected scenarios describing possible evolutions of the repository system in the post-closure phase, quantitative consequence analyses are performed. Due to the complexity of the phenomena of concern and the large timeframes under consideration, several types of uncertainties have to be taken into account. The modelling work for the far-field (geosphere) surrounding or overlaying the repository is based on model calculations concerning the groundwater movement and the resulting migration of radionuclides which possibly will be released from the repository. In contrast to engineered systems, the geosphere shows a strong spatial variability of facies, materials and material properties. The paper presented here describes the first steps towards a quantitative approach for an uncertainty assessment taking into account this variability. Due to the availability of a large amount of data and information of several types, the Gorleben site (Germany) has been used for a case study in order to demonstrate the method. (orig.)

  18. Regulatory support activities of JNES by thermal-hydraulic and safety analyses

    Kasahara, Fumio

    2008-01-01

    Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

  19. Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

    You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

    2004-10-01

    The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

  20. System Guidelines for EMC Safety-Critical Circuits: Design, Selection, and Margin Demonstration

    Lawton, R. M.

    1996-01-01

    Demonstration of safety margins for critical points (circuits) has traditionally been required since it first became a part of systems-level Electromagnetic Compatibility (EMC) requirements of MIL-E-6051C. The goal of this document is to present cost-effective guidelines for ensuring adequate Electromagnetic Effects (EME) safety margins on spacecraft critical circuits. It is for the use of NASA and other government agencies and their contractors to prevent loss of life, loss of spacecraft, or unacceptable degradation. This document provides practical definition and treatment guidance to contain costs within affordable limits.

  1. Compliance demonstration: What can be reasonably expected from safety assessment for geological repositories?

    Zuidema, P.; Smith, P.; Sumerling, T.

    1999-01-01

    When licensing a nuclear facility, it is important to demonstrate that it will comply with regulatory limits (e.g. individual dose limits) and also show that sufficient attention has been paid to optimisation of facility design and operation, such that any associated radiological impacts will be as low as reasonably achievable (ALARA). In general, in demonstrating compliance, experience can be drawn from the performance of existing and similar facilities, and monitoring plans can be specified that will confirm that actual radiological discharges during operations are within authorised limits for the facility. This is also true in respect of the operational period of a geological repository. For the post-closure phase of a repository, however, it is also necessary to show that possible releases will remain acceptably low even at long times in the future when, it is assumed, control of the facility has lapsed and there is no method of either monitoring releases or taking remedial action in the case of unexpected events or releases. In addition, within each country, a deep geological repository will be a first-of-a-kind development so that compliance arguments can be expected to be rigorously tested without any assistance from the precedent of licensing of similar facilities nationally. This puts heavy, and quite unusual, burdens on the long-term safety assessment for a geological repository to develop a case that is sufficiently strong to demonstrate compliance. This paper focuses on the problem of demonstrating compliance with long-term safety requirements for a geological repository, and explores: the overall aims and special difficulties of demonstrating compliance for a geological repository; the role of safety assessment in demonstrating compliance; the scope for optimisation of a geological repository and importance of robustness and lessons learnt from the application of safety assessment. In addition, some issues requiring further discussion and clarification

  2. Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

    1983-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches

  3. Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

    1983-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches.

  4. Detailed energy saving performance analyses on thermal mass walls demonstrated in a zero energy house

    Zhu, L. [School of Architecture, Tianjin University, Tianjin 300072 (China); Hurt, R.; Correia, D.; Boehm, R. [Center for Energy Research, University of Nevada, Las Vegas, NV 89154 (United States)

    2009-03-15

    An insulated concrete wall system{sup 1}1 was used on exterior walls of a zero energy house. Its thermal functions were investigated using actual data in comparison to a conventional wood frame system. The internal wall temperature of massive systems changes more slowly than the conventional wall constructions, leading to a more stable indoor temperature. The Energy10 simulated equivalent R-value and DBMS of the mass walls under actual climate conditions are, respectively, 6.98 (m{sup 2} C)/W and 3.39. However, the simulated heating energy use was much lower for the massive walls while the cooling load was a little higher. Further investigation on the heat flux indicates that the heat actually is transferred inside all day and night, which results in a higher cooling energy consumption. A one-dimensional model further verified these analyses, and the calculated results are in good agreement with the actual data. We conclude that the thermal mass wall does have the ability to store heat during the daytime and release it back at night, but in desert climates with high 24-h ambient temperature and intense sunlight, more heat will be stored than can be transferred back outside at night. As a result, an increased cooling energy will be required. (author)

  5. Demonstrating the efficiency of the EFPC criterion by means of Sensitivity analyses

    Munier, Raymond

    2007-04-01

    Within the framework of a project to characterise large fractures, a modelling effort was initiated to evaluate the use of a pair of full perimeter criteria, FPC and EFPC, for detecting fractures that could jeopardize the integrity of the canisters in the case of a large nearby earthquake. Though some sensitivity studies were performed in the method study of these mainly targeted aspects of Monte-Carlo simulations. The impact of uncertainties in the DFN model upon the efficiency of the FPI criteria was left unattended. The main purpose of this report is, therefore, to explore the impact of DFN variability upon the efficiency of the FPI criteria. The outcome of the present report may thus be regarded as complementary analyses to the ones presented in SKB-R-06-54. To appreciate the details of the present report, the reader should be acquainted with the simulation procedure described the earlier report. The most important conclusion of this study is that the efficiency of the EFPC is high for all tested model variants. That is, compared to blind deposition, the EFPC is a very powerful tool to identify unsuitable deposition holes and it is essentially insensitive to variations in the DFN Model. If information from adjacent tunnels is used in addition to EFPC, then the probability of detecting a critical deposition hole is almost 100%

  6. Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

    Sink, D.A.; Gibson, G.

    1979-03-01

    The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based on two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma

  7. Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety

    Ammerman, D.J.

    1997-01-01

    Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A 2 quantities). These packages have the most stringent performance criteria. Material with less than an A 2 quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in open-quotes strong-tightclose quotes packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material

  8. Code development and analyses within the area of transmutation and safety

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  9. Results and implications of the EBR-II inherent safety demonstration tests

    Planchon, H.P.; Golden, G.H.; Sackett, J.I.; Mohr, D.; Chang, L.K.; Feldman, E.E.; Betten, P.R.

    1987-01-01

    On April 3, 1986 two milestone tests were conducted in Experimental Breeder Reactor-2 (EBR-II). The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power and in both tests the reactor was shut down by natural processes, principally thermal expansion, without automatic scram, operator intervention or the help of special in-core devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. In a general sense, therefore, the tests plus supporting analysis demonstrated the feasibility of inherent passive shutdown for undercooling accidents in metal-fueled LMRs. The results provide a technical basis for future experiments in EBR-II to demonstrate inherent safety for overpower accidents and provide data for validation of computer codes used for design and safety analysis of inherently safe reactor plants

  10. Environment, Safety, Health, and Quality Plan for the Buried Waste Integrated Demonstration Program

    Walker, S.

    1994-05-01

    The Buried Waste Integrated Demonstration (BWID) is a program funded by the US Department of Energy Office of Technology Development. BWID supports the applied research, development, demonstration, testing, and evaluation of a suite of advanced technologies that together form a comprehensive remediation system for the effective and efficient remediation of buried waste. This document describes the Environment, Safety, Health, and Quality requirements for conducting BWID activities at the Idaho National Engineering Laboratory. Topics discussed in this report, as they apply to BWID operations, include Federal, State of Idaho, and Environmental Protection Agency regulations, Health and Safety Plans, Quality Program Plans, Data Quality Objectives, and training and job hazard analysis. Finally, a discussion is given on CERCLA criteria and System and Performance audits as they apply to the BWID Program

  11. Relocation work of temporary thermocouples for measuring the vessel cooling system in the safety demonstration test

    Shimazaki, Yosuke; Shinohara, Masanori; Ono, Masato; Yanagi, Shunki; Tochio, Daisuke; Iigaki, Kazuhiko

    2012-05-01

    It is necessary to confirm that the temperature of water cooling panel of the vessel cooling system (VCS) is controlled under the allowable working temperature during the safety demonstration test because the water cooling panel temperature rises due to stop of cooling water circulation pumps. Therefore, several temporary thermocouples are relocated to the water cooling panel near the stabilizers of RPV and the side cooling panel outlet ring header of VCS in order to observe the temperature change of VCS. The relocated thermocouples can measure the temperature change with starting of the cooling water circulation pumps of VCS. So it is confirmed that the relocated thermocouples can observe the VCS temperature change in the safety demonstration test. (author)

  12. Comparative Genome Analyses of Streptococcus suis Isolates from Endocarditis Demonstrate Persistence of Dual Phenotypic Clones.

    Tohya, Mari; Watanabe, Takayasu; Maruyama, Fumito; Arai, Sakura; Ota, Atsushi; Athey, Taryn B T; Fittipaldi, Nahuel; Nakagawa, Ichiro; Sekizaki, Tsutomu

    2016-01-01

    Many bacterial species coexist in the same niche as heterogeneous clones with different phenotypes; however, understanding of infectious diseases by polyphenotypic bacteria is still limited. In the present study, encapsulation in isolates of the porcine pathogen Streptococcus suis from persistent endocarditis lesions was examined. Coexistence of both encapsulated and unencapsulated S. suis isolates was found in 26 out of 59 endocarditis samples. The isolates were serotype 2, and belonged to two different sequence types (STs), ST1 and ST28. The genomes of each of the 26 pairs of encapsulated and unencapsulated isolates from the 26 samples were sequenced. The data showed that each pair of isolates had one or more unique nonsynonymous mutations in the cps gene, and the encapsulated and unencapsulated isolates from the same samples were closest to each other. Pairwise comparisons of the sequences of cps genes in 7 pairs of encapsulated and unencapsulated isolates identified insertion/deletions (indels) ranging from one to 104 bp in different cps genes of unencapsulated isolates. Capsule expression was restored in a subset of unencapsulated isolates by complementation in trans with cps expression vectors. Examination of gene content common to isolates indicated that mutation frequency was higher in ST28 pairs than in ST1 pairs. Genes within mobile genetic elements were mutation hot spots among ST28 isolates. Taken all together, our results demonstrate the coexistence of dual phenotype (encapsulated and unencapsulated) bacterial clones and suggest that the dual phenotypes arose independently in each farm by means of spontaneous mutations in cps genes.

  13. Demonstrated operational and inherent safety of the prototype fast reactor (PFR)

    Smedley, J.A.; Gregory, C.V.; Judd, A.M.

    1983-01-01

    The Prototype Fast Reactor (PFR) is sited at Dounreay, on the north coast of Scotland in the United Kingdom, and has been in operation since 1974. Three aspects of the safety of the reactor are described, including the all-important practical consideration of operational safety, a demonstration of the limited consequences of a sodium/water reaction in a steam generator and the ability of the reactor to protect itself against highly improbable incidents. Attention is drawn to the low radiation levels in the plant and the correspondingly low dose rate to personnel. A feature of PFR operation has been the stable and predictable behaviour of its core together with the high degree of reliability exhibited by the engineered safety system. No failures have occurred within the standard driver charge but two experimental fuel pins suffered cladding failure, which was detected easily by the fission gas and delayed neutron detection systems. In the steam generating units sodium and water are separated by the single steel wall of the steam tubes. Although no under-sodium leak has occurred, an experimental programme is continuing and demonstrates that were any such leak to occur its consequences would be containable and would not result in the release of sodium to the environment or any breach of the reactor containment. The final section describes the inherent safety features of the reactor which enable it to survive a range of very improbable incidents even when the engineered safeguards fail. The features considered are natural circulation, which has been demonstrated by reactor experiment; the reactor's negative power coefficient, which, for example, enables the reactor to survive a complete loss of heat sink; and the durability of the fuel pins, demonstrated by a series of boiling experiments in the Dounreay Fast Reactor (DFR). (author)

  14. Comparative Genome Analyses of Streptococcus suis Isolates from Endocarditis Demonstrate Persistence of Dual Phenotypic Clones.

    Mari Tohya

    Full Text Available Many bacterial species coexist in the same niche as heterogeneous clones with different phenotypes; however, understanding of infectious diseases by polyphenotypic bacteria is still limited. In the present study, encapsulation in isolates of the porcine pathogen Streptococcus suis from persistent endocarditis lesions was examined. Coexistence of both encapsulated and unencapsulated S. suis isolates was found in 26 out of 59 endocarditis samples. The isolates were serotype 2, and belonged to two different sequence types (STs, ST1 and ST28. The genomes of each of the 26 pairs of encapsulated and unencapsulated isolates from the 26 samples were sequenced. The data showed that each pair of isolates had one or more unique nonsynonymous mutations in the cps gene, and the encapsulated and unencapsulated isolates from the same samples were closest to each other. Pairwise comparisons of the sequences of cps genes in 7 pairs of encapsulated and unencapsulated isolates identified insertion/deletions (indels ranging from one to 104 bp in different cps genes of unencapsulated isolates. Capsule expression was restored in a subset of unencapsulated isolates by complementation in trans with cps expression vectors. Examination of gene content common to isolates indicated that mutation frequency was higher in ST28 pairs than in ST1 pairs. Genes within mobile genetic elements were mutation hot spots among ST28 isolates. Taken all together, our results demonstrate the coexistence of dual phenotype (encapsulated and unencapsulated bacterial clones and suggest that the dual phenotypes arose independently in each farm by means of spontaneous mutations in cps genes.

  15. Comparative Genome Analyses of Streptococcus suis Isolates from Endocarditis Demonstrate Persistence of Dual Phenotypic Clones

    Tohya, Mari; Watanabe, Takayasu; Maruyama, Fumito; Arai, Sakura; Ota, Atsushi; Athey, Taryn B. T.; Fittipaldi, Nahuel; Nakagawa, Ichiro; Sekizaki, Tsutomu

    2016-01-01

    Many bacterial species coexist in the same niche as heterogeneous clones with different phenotypes; however, understanding of infectious diseases by polyphenotypic bacteria is still limited. In the present study, encapsulation in isolates of the porcine pathogen Streptococcus suis from persistent endocarditis lesions was examined. Coexistence of both encapsulated and unencapsulated S. suis isolates was found in 26 out of 59 endocarditis samples. The isolates were serotype 2, and belonged to two different sequence types (STs), ST1 and ST28. The genomes of each of the 26 pairs of encapsulated and unencapsulated isolates from the 26 samples were sequenced. The data showed that each pair of isolates had one or more unique nonsynonymous mutations in the cps gene, and the encapsulated and unencapsulated isolates from the same samples were closest to each other. Pairwise comparisons of the sequences of cps genes in 7 pairs of encapsulated and unencapsulated isolates identified insertion/deletions (indels) ranging from one to 104 bp in different cps genes of unencapsulated isolates. Capsule expression was restored in a subset of unencapsulated isolates by complementation in trans with cps expression vectors. Examination of gene content common to isolates indicated that mutation frequency was higher in ST28 pairs than in ST1 pairs. Genes within mobile genetic elements were mutation hot spots among ST28 isolates. Taken all together, our results demonstrate the coexistence of dual phenotype (encapsulated and unencapsulated) bacterial clones and suggest that the dual phenotypes arose independently in each farm by means of spontaneous mutations in cps genes. PMID:27433935

  16. Genetic and functional analyses demonstrate a role for abnormal glycinergic signaling in autism.

    Pilorge, M; Fassier, C; Le Corronc, H; Potey, A; Bai, J; De Gois, S; Delaby, E; Assouline, B; Guinchat, V; Devillard, F; Delorme, R; Nygren, G; Råstam, M; Meier, J C; Otani, S; Cheval, H; James, V M; Topf, M; Dear, T N; Gillberg, C; Leboyer, M; Giros, B; Gautron, S; Hazan, J; Harvey, R J; Legendre, P; Betancur, C

    2016-07-01

    Autism spectrum disorder (ASD) is a common neurodevelopmental condition characterized by marked genetic heterogeneity. Recent studies of rare structural and sequence variants have identified hundreds of loci involved in ASD, but our knowledge of the overall genetic architecture and the underlying pathophysiological mechanisms remains incomplete. Glycine receptors (GlyRs) are ligand-gated chloride channels that mediate inhibitory neurotransmission in the adult nervous system but exert an excitatory action in immature neurons. GlyRs containing the α2 subunit are highly expressed in the embryonic brain, where they promote cortical interneuron migration and the generation of excitatory projection neurons. We previously identified a rare microdeletion of the X-linked gene GLRA2, encoding the GlyR α2 subunit, in a boy with autism. The microdeletion removes the terminal exons of the gene (GLRA2(Δex8-9)). Here, we sequenced 400 males with ASD and identified one de novo missense mutation, p.R153Q, absent from controls. In vitro functional analysis demonstrated that the GLRA2(Δex8)(-)(9) protein failed to localize to the cell membrane, while the R153Q mutation impaired surface expression and markedly reduced sensitivity to glycine. Very recently, an additional de novo missense mutation (p.N136S) was reported in a boy with ASD, and we show that this mutation also reduced cell-surface expression and glycine sensitivity. Targeted glra2 knockdown in zebrafish induced severe axon-branching defects, rescued by injection of wild type but not GLRA2(Δex8-9) or R153Q transcripts, providing further evidence for their loss-of-function effect. Glra2 knockout mice exhibited deficits in object recognition memory and impaired long-term potentiation in the prefrontal cortex. Taken together, these results implicate GLRA2 in non-syndromic ASD, unveil a novel role for GLRA2 in synaptic plasticity and learning and memory, and link altered glycinergic signaling to social and cognitive

  17. Swiss-Slovak cooperation program: a training strategy for safety analyses

    Husarcek, J.

    2000-01-01

    During the 1996-1999 period, a new training strategy for safety analyses was implemented at the Slovak Nuclear Regulatory Authority (UJD) within the Swiss-Slovak cooperation programme in nuclear safety (SWISSLOVAK). The SWISSLOVAK project involved the recruitment, training, and integration of the newly established team into UJD's organizational structure. The training strategy consisted primarily of the following two elements: a) Probabilistic Safety Analysis (PSA) applications (regulatory review and technical evaluation of Level-1/Level-2 PSAs; PSA-based operational events analysis, PSA applications to assessment of Technical Specifications; and PSA-based hardware and/or procedure modifications) and b) Deterministic accident analyses (analysis of accidents and regulatory review of licensee Safety Analysis Reports; analysis of severe accidents/radiological releases and the potential impact of the containment and engineered safety systems, including the development of technical bases for emergency response planning; and application of deterministic methods for evaluation of accident management strategies/procedure modifications). The paper discusses the specific aspects of the training strategy performed at UJD in both the probabilistic and deterministic areas. The integration of team into UJD's organizational structure is described and examples of contributions of the team to UJD's statutory responsibilities are provided. (author)

  18. Demonstration testing and evaluation of in situ soil heating. Health and safety plan (Revision 2)

    Dev, H.

    1994-12-28

    This document is the Health and Safety Plan (HASP) for the demonstration of IITRI`s EM Treatment Technology. In this process, soil is heated in situ by means of electrical energy for the removal of hazardous organic contaminants. This process will be demonstrated on a small plot of contaminated soil located in the Pit Area of Classified Burial Ground K-1070-D, K-25 Site, Oak Ridge, TN. The purpose of the demonstration is to remove organic contaminants present in the soil by heating to a temperature range of 85{degrees} to 95{degrees}C. The soil will be heated in situ by applying 60-Hz AC power to an array of electrodes placed in boreholes drilled through the soil. In this section a brief description of the process is given along with a description of the site and a listing of the contaminants found in the area.

  19. Demonstration testing and evaluation of in situ soil heating. Health and safety plan (Revision 2)

    Dev, H.

    1994-01-01

    This document is the Health and Safety Plan (HASP) for the demonstration of IITRI's EM Treatment Technology. In this process, soil is heated in situ by means of electrical energy for the removal of hazardous organic contaminants. This process will be demonstrated on a small plot of contaminated soil located in the Pit Area of Classified Burial Ground K-1070-D, K-25 Site, Oak Ridge, TN. The purpose of the demonstration is to remove organic contaminants present in the soil by heating to a temperature range of 85 degrees to 95 degrees C. The soil will be heated in situ by applying 60-Hz AC power to an array of electrodes placed in boreholes drilled through the soil. In this section a brief description of the process is given along with a description of the site and a listing of the contaminants found in the area

  20. Safety analyses of the nuclear-powered ship Mutsu with RETRAN

    Naruko, Y.; Ishida, T.; Tanaka, Y.; Futamura, Y.

    1982-01-01

    To provide a quantitative basis for the safety evaluation of the N.S. Mutsu, a number of safety analyses were performed in the course of reexamination. With respect to operational transient analyses, the RETRAN computer code was used to predict plant performances on the basis of postulated transient scenarios. The COBRA-IV computer code was also used to obtain a value of the minimum DNBR for each transient, which is necessary to predict detailed thermal-hydraulic performances in the core region of the reactor. In the present paper, the following three operational transients, which were calculated as a part of the safety analyses, are being dealt with: a complete loss of load without reactor scram; an excessive load increase incident, which is followed by a 30 percent stepwise load increase in the steam dump flow; and an accidental depressurization of the primary system, which is followed by a sudden full opening of the pressurizer spray valve. A Mutsu two-loop RETRAN model and simulation results were described. The results being compared with those of land-based PWRs, the characteristic features of the Mutsu reactor were presented and the safety of the plant under the operational transient conditions was confirmed

  1. Sensitivity and uncertainty analyses applied to criticality safety validation. Volume 2

    Broadhead, B.L.; Hopper, C.M.; Parks, C.V.

    1999-01-01

    This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies developed in Volume 1 to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the existing S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently in use by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The methods for application of S/U and generalized linear-least-square methodology (GLLSM) tools to the criticality safety validation procedures were described in Volume 1 of this report. Volume 2 of this report presents the application of these procedures to the validation of criticality safety analyses supporting uranium operations where enrichments are greater than 5 wt %. Specifically, the traditional k eff trending analyses are compared with newly developed k eff trending procedures, utilizing the D and c k coefficients described in Volume 1. These newly developed procedures are applied to a family of postulated systems involving U(11)O 2 fuel, with H/X values ranging from 0--1,000. These analyses produced a series of guidance and recommendations for the general usage of these various techniques. Recommendations for future work are also detailed

  2. Process hazards analysis (PrHA) program, bridging accident analyses and operational safety

    Richardson, J.A.; McKernan, S.A.; Vigil, M.J.

    2003-01-01

    Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker

  3. European passive plant program preliminary safety analyses to support system design

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

    1999-01-01

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

  4. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  5. Analysing context-dependent deviations in interacting with safety-critical systems

    Paterno, Fabio; Santoro, Carmen

    2006-01-01

    Mobile technology is penetrating many areas of human life. This implies that the context of use can vary in many respects. We present a method that aims to support designers in managing the complex design space when considering applications with varying contexts and help them to identify solutions that support users in performing their activities while preserving usability and safety. The method is a novel combination of an analysis of both potential deviations in task performance and most suitable information representations based on distributed cognition. The originality of the contribution is in providing a conceptual tool for better understanding the impact of context of use on user interaction in safety-critical domains. In order to present our approach we provide an example in which the implications of introducing new support through mobile devices in a safety-critical system are identified and analysed in terms of potential hazards

  6. Current regulatory developments concerning the implementation of probabilistic safety analyses for external hazards in Germany

    Krauss, Matias; Berg, Heinz-Peter

    2014-01-01

    The Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) initiated in September 2003 a comprehensive program for the revision of the national nuclear safety regulations which has been successfully completed in November 2012. These nuclear regulations take into account the current recommendations of the International Atomic Energy Agency (IAEA) and Western European Nuclear Regulators Association (WENRA). In this context, the recommendations and guidelines of the Nuclear Safety Standards Commission (KTA) and the technical documents elaborated by the respective expert group on Probabilistic Safety Analysis for Nuclear Power Plants (FAK PSA) are being updated or in the final process of completion. A main topic of the revision was the issue external hazards. As part of this process and in the light of the accident at Fukushima and the findings of the related actions resulting in safety reviews of nuclear power plants at national level in Germany and on European level, a revision of all relevant standards and documents has been made, especially the recommendations of KTA and FAK PSA. In that context, not only design issues with respect to events such as earthquakes and floods have been discussed, but also methodological issues regarding the implementation of improved probabilistic safety analyses on this topic. As a result of the revision of the KTA 2201 series 'Design of Nuclear Power Plants against Seismic Events' with their parts 1 to 6, part 1 'Principles' was published as the first standard in November 2011, followed by the revised versions of KTA 2201.2 (soil) and 2201.4 (systems and components) in 2012. The modified the standard KTA 2201.3 (structures) is expected to be issued before the end of 2013. In case of part 5 (seismic instrumentation) and part 6 (post>seismic actions) draft amendments are expected in 2013. The expert group 'Probabilistic Safety Assessments for Nuclear Power Plants' (FAK PSA) is an advisory body of the Federal

  7. Dry critical experiments and analyses performed in support of the Topaz-2 Safety Program

    Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Loynstev, V.A.

    1994-01-01

    In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations

  8. Safety and deterministic failure analyses in high-beta D-D tokamak reactors

    Selcow, E.C.

    1984-01-01

    Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

  9. C4P cross-section libraries for safety analyses with SIMMER and related studies

    Rineiski, A.; Sinitsa, V.; Gabrielli, F.; Maschek, W.

    2011-01-01

    A code and data system, C 4 P, is under development at KIT. It includes fine-group master libraries and tools for generating problem-oriented cross-section libraries, primarily for safety studies with the SIMMER code and related analyses. In the paper, the 560-group master library and problem oriented 40-group and 72-group cross-section libraries, for thermal and fast systems, respectively, are described and their performances are investigated. (author)

  10. Selected problems and results of the transient event and reliability analyses for the German safety study

    Hoertner, H.

    1977-01-01

    For the investigation of the risk of nuclear power plants loss-of-coolant accidents and transients have to be analyzed. The different functions of the engineered safety features installed to cope with transients are explained. The event tree analysis is carried out for the important transient 'loss of normal onsite power'. Preliminary results of the reliability analyses performed for quantitative evaluation of this event tree are shown. (orig.) [de

  11. Demonstrating safety during license renewal should not be a large task

    Berto, D.S.

    1993-01-01

    The principal regulatory goal related to nuclear power plant operation is to ensure the health and safety of the public. The principal goal of extended plant operation via the license renewal process is also to ensure the health and safety of the public. The license renewal documentation issued by the Nuclear Regulatory Commission (NRC) provides guidance on what will be acceptable to the NRC in a license renewal application to demonstrate that this goal will be met. Application of this guidance is currently open to wide interpretation, with many of the current approaches proving to be extremely costly, complex, and uncertain of acceptability. This paper evaluates the requirements necessary to ensure the continued health and safety of the public during any license renewal term. This evaluation is based on the stated goals of the License Renewal Rule and on the published bases for the Rule. An approach to License Renewal is recommended that: (1) meets the stated goals of the NRC; (2) is consistent with current regulatory practices; and (3) will continue to ensure the health and safety of the public. This recommended approach is also much less costly than other current approaches, and can be easily agreed to by all participants. This approach will meet regulatory goals, while removing the cost and uncertainty obstacles currently being confronted by utilities. Providing a viable approach to license renewal will allow the renewal process to be pursued by utilities. Without such an approach, safe and reliable nuclear power plants will be permanently shut down at the arbitrary 40 year license limit

  12. Response to 'Audiences, rationales and quantitative measure for demonstrations of nuclear safety and licensing by tests'

    Taylor, J J [Electric Power Research Institute, Palo Alto, CA (United States)

    1990-07-01

    There are key overriding issues which are independent of the specific nature of the nuclear system itself which require concentrated attention to assure public safety and reliable, economic operation: - the need to keep the risk of external events to an acceptable level for all reactor systems; - the need to assure highly reliable operation of all elements of the system, many of which are the same regardless of what the nuclear system is composed of; - the importance of human proficiency in operating this total complex in a highly reliable manner. Nuclear system-specific demonstrations of public safety, although valuable, will not accomplish this and will not convince the public that there is zero risk. The very claim that a nuclear system or for that matter any big industrial complex, poses zero public risk raises a credibility gap with the public and is, therefore, counterproductive. So, we must take the dull, detailed technical steps to address the challenge: - define the minimal risk and accept that there is no zero risk; - demonstrate the achievement of that risk by detailed testing, conformance to standards and regulation, and trouble-free operation.

  13. Response to 'Audiences, rationales and quantitative measure for demonstrations of nuclear safety and licensing by tests'

    Taylor, J.J.

    1990-01-01

    There are key overriding issues which are independent of the specific nature of the nuclear system itself which require concentrated attention to assure public safety and reliable, economic operation: - the need to keep the risk of external events to an acceptable level for all reactor systems; - the need to assure highly reliable operation of all elements of the system, many of which are the same regardless of what the nuclear system is composed of; - the importance of human proficiency in operating this total complex in a highly reliable manner. Nuclear system-specific demonstrations of public safety, although valuable, will not accomplish this and will not convince the public that there is zero risk. The very claim that a nuclear system or for that matter any big industrial complex, poses zero public risk raises a credibility gap with the public and is, therefore, counterproductive. So, we must take the dull, detailed technical steps to address the challenge: - define the minimal risk and accept that there is no zero risk; - demonstrate the achievement of that risk by detailed testing, conformance to standards and regulation, and trouble-free operation

  14. Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

    Bollingerfehr, W.; Buhmann, D.; Doerr, S.; and others

    2017-03-15

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  15. Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

    Bollingerfehr, W.; Buhmann, D.; Doerr, S.

    2017-03-01

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  16. Bavarian liquid hydrogen bus demonstration project - safety, licensing and acceptability aspects

    Wurster, R.; Knorr, H.; Pruemm, W.

    1999-07-01

    A regular 12 m city bus of the MAN SL 202 type with an internal combustion engine adapted to hydrogen operation and auxiliary gasoline operation was demonstrated in the Bavarian cities of Erlangen and Munich between April 1996 and August 1998. Three bus operators, Erlanger Stadtwerke, Stadtwerke Muenchen and Autobus Oberbayern were testing the bus in three different operating schemes. In order to be able to perform this worldwide first public demonstration of a liquid hydrogen (LH{sub 2}) city bus in regular service, several requirements with respect to safety, licensing, training and acceptability had to be fulfilled. These activities were focusing mainly on the hydrogen specific issues such as (a) integration of onboard LH{sub 2} storage vessels, piping and instrumentation, (b) implementation of storage and refueling infrastructure in the operators' yards, (c) adaptation of the maintenance garages, (d) training of operating and maintenance personnel. During phase II of the demonstration activity a poll was performed on passengers traveling onboard the hydrogen-powered city bus in order to determined the level of acceptance among the users of the bus. The bus was designed and manufactured by MAN Nutzfahrzeuge Aktiengesellschaft. The cryogenic fuel storage and the refueling equipment were designed and manufactured by Linde AG. The realization of the hardware was financially supported by the European Commission (EC) within the Euro-Quebec Hydro-Hydrogen Pilot Project. The demonstration phase was financially supported by EC and the Bavarian State Government. Ludwig-Boelkow-Systemtechnik performed project monitoring for both funding organizations. The presentation will summarize the most important results of this demonstration phase and will address the measures undertaken in order to get the bus, the refueling infrastructure and the maintenance and operating procedures approved by the relevant authorities.

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

    Elvik, Rune

    2003-09-01

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

  18. Safety culture and learning from incidents: the role of incident reporting and causal analyses

    Wilpert, B.

    1994-01-01

    Nuclear industry more than any other industrial branch has developed and used predictive risk analysis as a method of feedforward control of safety and reliability. Systematic evaluation of operating experience, statistical documentation of component failures, systematic documentation and analysis of incidents are important complementary elements of feedback control: we are dealing here with adjustment and learning from experience, in particular from past incidents. Using preliminary findings from ongoing research at the Research Center Systems Safety at the Berlin University of Technology the contribution discusses preconditions for an effective use of lessons to be learnt from closely matched incident reporting and in depth analyses of causal chains leading to incidents. Such conditions are especially standardized documentation, reporting and analyzing methods of incidents; structured information flows and feedback loops; abstaining from culpability search; mutual trust of employees and management; willingness of all concerned to continually evaluate and optimize the established learning system. Thus, incident related reporting and causal analyses contribute to safety culture, which is seen to emerge from tightly coupled organizational measures and respective change in attitudes and behaviour. (author) 2 figs., 7 refs

  19. Scoping analyses for the safety injection system configuration for Korean next generation reactor

    Bae, Kyoo Hwan; Song, Jin Ho; Park, Jong Kyoon

    1996-01-01

    Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are performed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSl pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SlT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA

  20. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  1. Technical and administrative approach for the West Valley Demonstration Project Safety Program

    Newsom, P.C.; Roberts, C.J.; Yuchien Yuan; Marchetti, S.

    1987-06-01

    The principal objective of the West Valley Demonstration Project (WVDP) is to vitrify the 2.2 million liters of high-level radioactive waste (HLW) stored at the Western New York Nuclear Service Center (WNYNSC). This simple statement of purpose, however, does not convey a sense of the complexity of the undertaking. The vitrification task is not only complex in and of itself, but requires a myriad of other activities to be accomplished on an intricate and fast paced schedule in order to support it. The West Valley Demonstration Project Act (P.L 96-368), U.S. Department of Energy Order DOE-5481.1A, Idaho Operations Office Order ID-5481.1 and standard nuclear industry practice all require that proposed systems and operations involving hazards not routinely encountered by the general public be analyzed to identify potential hazards and consequences, and to assure that reasonable measures are taken to eliminate, control, or mitigate these potential consequences. Virtually every substantive aspect of the WVDP involves hazards beyond those routinely encountered and accepted by the general public. In order to assure the safety of the public and the workers at the WVDP, a system of hazard identification, categorization, analysis and review has been established. In parallel with this system, a procedure for developing the minimum design specifications and quality assurance requirements has been developed for Project systems, components, and structures which play a role in the safety of a specific major facility or the overall Project. 29 refs., 3 figs., 6 tabs

  2. Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  3. Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

    Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

    1984-08-01

    JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

  4. Safety analyses for an in-pile SCWR fuel qualification test loop

    Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

    2014-07-01

    As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

  5. Safety and sensitivity analyses of a generic geologic disposal system for high-level radioactive waste

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

    1994-11-01

    This report describes safety and sensitivity analyses of a generic geologic disposal system for HLW, using a GSRW code and an automated sensitivity analysis methodology based on the Differential Algebra. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. The results of sensitivity analyses indicate that parameters related to a homogeneous rock surrounding a disposal facility have higher sensitivities to the output analyzed here than those of a fractured zone and engineered barriers. The sensitivity analysis methodology provides technical information which might be bases for the optimization of design of the disposal facility. Safety analyses were performed on the reference disposal system which involve HLW in amounts corresponding to 16,000 MTU of spent fuels. The individual dose equivalent due to the exposure pathway ingesting drinking water was calculated using both the conservative and realistic values of geochemical parameters. In both cases, the committed dose equivalent evaluated here is the order of 10 -7 Sv, and thus geologic disposal of HLW may be feasible if the disposal conditions assumed here remain unchanged throughout the periods assessed here. (author)

  6. Recognising safety critical events: can automatic video processing improve naturalistic data analyses?

    Dozza, Marco; González, Nieves Pañeda

    2013-11-01

    New trends in research on traffic accidents include Naturalistic Driving Studies (NDS). NDS are based on large scale data collection of driver, vehicle, and environment information in real world. NDS data sets have proven to be extremely valuable for the analysis of safety critical events such as crashes and near crashes. However, finding safety critical events in NDS data is often difficult and time consuming. Safety critical events are currently identified using kinematic triggers, for instance searching for deceleration below a certain threshold signifying harsh braking. Due to the low sensitivity and specificity of this filtering procedure, manual review of video data is currently necessary to decide whether the events identified by the triggers are actually safety critical. Such reviewing procedure is based on subjective decisions, is expensive and time consuming, and often tedious for the analysts. Furthermore, since NDS data is exponentially growing over time, this reviewing procedure may not be viable anymore in the very near future. This study tested the hypothesis that automatic processing of driver video information could increase the correct classification of safety critical events from kinematic triggers in naturalistic driving data. Review of about 400 video sequences recorded from the events, collected by 100 Volvo cars in the euroFOT project, suggested that drivers' individual reaction may be the key to recognize safety critical events. In fact, whether an event is safety critical or not often depends on the individual driver. A few algorithms, able to automatically classify driver reaction from video data, have been compared. The results presented in this paper show that the state of the art subjective review procedures to identify safety critical events from NDS can benefit from automated objective video processing. In addition, this paper discusses the major challenges in making such video analysis viable for future NDS and new potential

  7. Scanning electron microscopic analyses of Ferrocyanide tank wastes for the Ferrocyanide safety program

    Callaway, W.S.

    1995-09-01

    This is Fiscal Year 1995 Annual Report on the progress of activities relating to the application of scanning electron microscopy in addressing the Ferrocyanide Safety Issue associated with Hanford Site high-level radioactive waste tanks. The status of the FY 1995 activities directed towards establishing facilities capable of providing SEM based micro-characterization of ferrocyanide tank wastes is described. A summary of key events in the SEM task over FY 1995 and target activities in FY 1996 are presented. A brief overview of the potential applications of computer controlled SEM analytical data in light of analyses of ferrocyanide simulants performed by an independent contractor is also presented

  8. International validation of safety analyses for nuclear power plants; Mednarodno preverjanje varnostnih analiz za jedrske elektrane

    Gregoric, N; Mavko, B [Institut ' Jozef Stefan' Ljubljana (Yugoslavia)

    1988-07-01

    Paper describes the participation of 'J.Stefan' Institute in international standard problems for validation of modeling and programs for safety analysis. Listed are main international experimental facilities for collecting data basic for understanding of physical phenomena, code development and validation of modelling and programs. Since the results of international standard problem analyses are published in a joint final report, it is simple to asses the conformance of the results of a particular group with the experiment. Good results from three international exercises done so far, have encouraged the group to currently participate in OECD-ISP-22 which is a model of the Italian three loop PWR. (author)

  9. Preliminary standard review guide for Environmental Restoration/Decontamination and Decommissioning safety analyses

    Ellingson, D.R.

    1993-06-01

    The review guide is based on the shared experiences, approaches, and philosophies of the Environmental Restoration/Decontamination and Decommissioning (ER/D ampersand D) subgroup members. It is presented in the form of a review guide to maximize the benefit to both the safety analyses practitioner and reviewer. The guide focuses on those challenges that tend to be unique to ER/D ampersand D cleanup activities. Some of these experiences, approaches, and philosophies may find application or be beneficial to a broader spectrum of activities such as terminal cleanout or even new operations. Challenges unique to ER/D ampersand D activities include (1) consent agreements requiring activity startup on designated dates; (2) the increased uncertainty of specific hazards; and (3) the highly variable activities covered under the broad category of ER/D ampersand D. These unique challenges are in addition to the challenges encountered in all activities; e.g., new and changing requirements and multiple interpretations. The experiences in approaches, methods, and solutions to the challenges are documented from the practitioner and reviewer's perspective, thereby providing the viewpoints on why a direction was taken and the concerns expressed. Site cleanup consent agreements with predetermined dates for restoration activity startup add the dimension of imposed punitive actions for failure to meet the date. Approval of the safety analysis is a prerequisite to startup. Actions that increase expediency are (1) assuring activity safety; (2) documenting that assurance; and (3) acquiring the necessary approvals. These actions increase the timeliness of startup and decrease the potential for punitive action. Improvement in expediency has been achieved by using safety analysis techniques to provide input to the line management decision process rather than as a review of line management decisions. Expediency is also improved by sharing the safety input and resultant decisions with

  10. Analyses to demonstrate the structural performance of the CASTOR KN12 in hypothetical accident drop accident scenarios

    Diersch, R.; Weiss, M.; Tso, C.F.; Chung, S.H.; Lee, H.Y.

    2004-01-01

    CASTORc ircledR KN-12 is a new cask design by GNB for KHNP-NETEC for dry and wet transportation of up to twelve spent PWR fuel assemblies in Korea. It received its transport license from the Korean Competent Authority KINS in July 2002 and is now in use in South Korea. It has been designed to satisfy the regulatory requirements of the 10 CFR 71 and the IAEA ST-1 for Type B(U)F packages. Its structural performance was demonstrated against the load cases and boundary conditions as defined in 10 CFR 71 and NRC's Regulatory Guide 7.8, and further explained in NUREG 1617. This included normal conditions of transport load cases - including Hot Environment, Cold Environment, Increased External Pressure (140MPa), Minimum External Pressure (24.5kPa), Vibration and shock, and 0.3m free drop - and the hypothetical accident conditions load cases - including the 9m Free Drop, Puncture, Thermal Fire Accident, 200m Water Immersion and 1.5 x MNOP Internal Pressure. Structural performance were demonstrated by analysis, including state-of-the-art finite element (FE) simulation, and confirmed by tests using a 1/3-scale model. Test results were also used to verify the numerical tool and the methods used in the analyses. All the structural analyses including validation against drop tests were carried out by Arup, and testing were carried out by KAERI. This paper concentrates on the analysis carried out to demonstrate performance in the hypothetical accident 9m free drop scenarios, and results from a small selection of them

  11. Analyses to demonstrate the structural performance of the CASTOR KN12 in hypothetical accident drop accident scenarios

    Diersch, R.; Weiss, M. [Gesellschaft fuer Nuklear-Behaelter mbH (Germany); Tso, C.F. [Arup (United Kingdom); Chung, S.H.; Lee, H.Y. [KHNP-NETEC (Korea)

    2004-07-01

    CASTORc{sup ircledR} KN-12 is a new cask design by GNB for KHNP-NETEC for dry and wet transportation of up to twelve spent PWR fuel assemblies in Korea. It received its transport license from the Korean Competent Authority KINS in July 2002 and is now in use in South Korea. It has been designed to satisfy the regulatory requirements of the 10 CFR 71 and the IAEA ST-1 for Type B(U)F packages. Its structural performance was demonstrated against the load cases and boundary conditions as defined in 10 CFR 71 and NRC's Regulatory Guide 7.8, and further explained in NUREG 1617. This included normal conditions of transport load cases - including Hot Environment, Cold Environment, Increased External Pressure (140MPa), Minimum External Pressure (24.5kPa), Vibration and shock, and 0.3m free drop - and the hypothetical accident conditions load cases - including the 9m Free Drop, Puncture, Thermal Fire Accident, 200m Water Immersion and 1.5 x MNOP Internal Pressure. Structural performance were demonstrated by analysis, including state-of-the-art finite element (FE) simulation, and confirmed by tests using a 1/3-scale model. Test results were also used to verify the numerical tool and the methods used in the analyses. All the structural analyses including validation against drop tests were carried out by Arup, and testing were carried out by KAERI. This paper concentrates on the analysis carried out to demonstrate performance in the hypothetical accident 9m free drop scenarios, and results from a small selection of them.

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

    1976-01-01

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

  13. Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

    Ruebel, Andre; Meleshyn, Artur

    2014-08-01

    The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

  14. Dry critical experiments and analyses performed in support of the TOPAZ-2 safety program

    Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Lobynstev, V.A.

    1995-01-01

    In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations. copyright 1995 American Institute of Physics

  15. Swiss regulatory use of databanks for nuclear power plant life management, surveillance and safety analyses

    Tipping, Ph.; Beutler, R.; Schoen, G.; Noeggerath, J.

    2002-01-01

    Full text: As operational time is accumulated, the overall safety and performance of nuclear power plants (NPPs) will tend to be characterised by those areas in which structures, systems and components (SSCs) have not performed as well, or as reliably, as expected. The reasons for non-availability of equipment in NPPs due to SSC material malfunction or unsatisfactory performance, leading to events or even accidents, are varied and they must be analysed in order to obtain the root causes. Once the root causes are identified, corresponding measures can be applied in order to improve reliability and therefore safety. The root cause information obtained, if brought into user-friendly databanks (DBs), can be used to follow NPP performance trends, to check whether a repair or replacement has been effective, to focus regulatory attention and NPP surveillance on known weak-spots and to serve as an advance indicator where potential problems may arise. Using the DBs, similar occurrences of failures or problems in other NPPs can be identified and generic issues recognised early on and preventative action taken. The following describes the Swiss Federal Nuclear Safety Inspectorate's (HSK) DB concepts for keeping track of NPP safety and lifetime management issues. Typical sources of data for the Inspectorate's DBs are, for example, the IAEA/NEA Incident Reporting System (IRS) reports, US-NRC Generic Letters, the Swiss NPP's own reports (monthly, annual and normal outage) and, more importantly, the document that these NPPs must issue to the Inspectorate whenever a reportable event takes place. Specifically, the reporting of events in the NPPs is laid down in the Inspectorate's Guideline (R-15 'Reporting Guideline Concerning The Operation of Nuclear Power Plants'). In this Guideline, reportable events are defined and the criteria for assessing the degree of importance or impact on nuclear safety are given. In this manner, a standard and consistent approach to data collection is

  16. International intercomparison and harmonization projects for demonstrating the safety of radioactive waste management, decommissioning and radioactive waste disposal

    Metcalf, Phil; O'Donnell, Patricio; Jova Sed, Luis; Batandjieva, Borislava; Rowat, John; Kinker, Monica

    2008-01-01

    Full text: The Joint Convention on the safety of spent fuel management and the safety of radioactive waste management and the international safety standards on radioactive waste management, decommissioning and radioactive waste disposal call for assessment and demonstration of the safety of facilities and activities; during siting, design and construction prior to operation, periodically during operation and at the end of lifetime or upon closure of a waste disposal facility. In addition, more recent revisions of the international safety standards require the development of a safety case for such facilities and activities, documentation presenting all the arguments supporting the safety of the facilities and activities covering site and engineering features, quantitative safety assessment and management systems. Guidance on meeting these safety requirements also indicates the need for a graded approach to safety assessment, with the extent and complexity of the assessment being proportional to the complexity of the activity or facility, and its propensity for radiation hazard. Safety assessment approaches and methodologies have evolved over several decades and international interest in these developments has been considerable as they can be complex and often subjective, which has led to international projects being established aimed at harmonization. The IAEA has sponsored a number of such initiatives, particularly in the area of disposal facility safety, but more recently in the areas of pre disposal waste management and decommissioning, including projects known as ISAM, ASAM, SADRWMS and DeSa. The projects have a number of common aspects including development of standardized methodological approaches, application on test cases and assessment review; they also have activity and facility specific elements. The paper presents an overview of the projects, the outcomes from the projects to date and their future direction aimed very much at practical application of

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

    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)

  18. Reentry safety for the Topaz II Space Reactor: Issues and analyses

    Connell, L.W.; Trost, L.C.

    1994-03-01

    This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents

  19. Time-frames and the demonstration of safety for HLW disposal

    Watkins, B.; Kessler, J.

    1999-01-01

    An important principle which is often embodied in the criteria for the safe disposal of long-lived radioactive wastes is that a similar level of radiation protection should be provided to future generations as that provided for those alive today. This has resulted in the development of performance assessment methodologies to evaluate the potential long term impacts of HLW disposal on humans, usually in terms of individual dose or risk. However, the actual periods of time over which it is expected that there will be full control over high level waste disposals are extremely short in comparison with the times over which radionuclides in the wastes could potentially move from the deep repository and emerge into the surface environment. This leads to problems in setting quantitative dose or risk based standard appropriate for the short and long term, and in setting the time-frames for which calculations should be carried out. This is especially difficult in view of the uncertainty in predicting changes in human behaviour and changes in the biosphere and geosphere over the time-scales involved. Different assessment time-frames and approaches proposed by IAEA, Nordic countries, Britain and US guidance documents are briefly reviewed. Whilst accepting the basic radiation protection objective of protecting future generations, no international consensus bas been agreed on what time-frames should be used in performance assessments. It is recommended that different time-frames should be associated with different quantitative or qualitative performance measures. As a result, a range of indicators of safety may be appropriate in demonstrating compliance with regulatory performance criteria and the consequent overall assessment context. It is argued that what is required is a simple, robust yet defensible approach to time-frames and performance indicators which can be accepted by the public, regulators and the nuclear industry

  20. Efficacy and Safety Extrapolation Analyses for Atomoxetine in Young Children with Attention-Deficit/Hyperactivity Disorder.

    Upadhyaya, Himanshu; Kratochvil, Christopher; Ghuman, Jaswinder; Camporeale, Angelo; Lipsius, Sarah; D'Souza, Deborah; Tanaka, Yoko

    2015-12-01

    This extrapolation analysis qualitatively compared the efficacy and safety profile of atomoxetine from Lilly clinical trial data in 6-7-year-old patients with attention-deficit/hyperactivity disorder (ADHD) with that of published literature in 4-5-year-old patients with ADHD (two open-label [4-5-year-old patients] and one placebo-controlled study [5-year-old patients]). The main efficacy analyses included placebo-controlled Lilly data and the placebo-controlled external study (5-year-old patients) data. The primary efficacy variables used in these studies were the ADHD Rating Scale-IV Parent Version, Investigator Administered (ADHD-RS-IV-Parent:Inv) total score, or the Swanson, Nolan and Pelham (SNAP-IV) scale score. Safety analyses included treatment-emergent adverse events (TEAEs) and vital signs. Descriptive statistics (means, percentages) are presented. Acute atomoxetine treatment improved core ADHD symptoms in both 6-7-year-old patients (n=565) and 5-year-old patients (n=37) (treatment effect: -10.16 and -7.42). In an analysis of placebo-controlled groups, the mean duration of exposure to atomoxetine was ∼ 7 weeks for 6-7-year-old patients and 9 weeks for 5-year-old patients. Decreased appetite was the most common TEAE in atomoxetine-treated patients. The TEAEs observed at a higher rate in 5-year-old versus 6-7-year-old patients were irritability (36.8% vs. 3.6%) and other mood-related events (6.9% each vs. atomoxetine may improve ADHD symptoms, but possibly to a lesser extent than in older children, with some adverse events occurring at a higher rate in 5-year-old patients.

  1. Safety analyses of potential exposure in medical irradiation plants by Fuzzy Fault Tree

    Casamirra, Maddalena; Castiglia, Francesco; Giardina, Mariarosa; Tomarchio, Elio

    2008-01-01

    The results of Fuzzy Fault Tree (FFT) analyses of various accidental scenarios, which involve the operators in potential exposures inside an High Dose Rate (HDR) remote after-loading systems for use in brachytherapy, are reported. To carry out fault tree analyses by means of fuzzy probabilities, the TREEZZY2 computer code is used. Moreover, the HEART (Human Error Assessment and Reduction Technique) model, properly modified on the basis of the fuzzy approach, has been employed to assess the impact of performances haping and error-promoting factors in the context of the accidental events. The assessment of potential dose values for some identified accidental scenarios allows to consider, for a particular event, a fuzzy uncertainty range in potential dose estimate. The availability of lower and upper limits allows evaluating the possibility of optimization of the installation from the point of view of radiation protection. The adequacy of the training and information program for staff and patients (and their family members) and the effectiveness of behavioural rules and safety procedures were tested. Some recommendations on procedures and equipment to reduce the risk of radiological exposure are also provided. (author)

  2. Japanese standard method for safety evaluation using best estimate code based on uncertainty and scaling analyses with statistical approach

    Mizokami, Shinya; Hotta, Akitoshi; Kudo, Yoshiro; Yonehara, Tadashi; Watada, Masayuki; Sakaba, Hiroshi

    2009-01-01

    Current licensing practice in Japan consists of using conservative boundary and initial conditions(BIC), assumptions and analytical codes. The safety analyses for licensing purpose are inherently deterministic. Therefore, conservative BIC and assumptions, such as single failure, must be employed for the analyses. However, using conservative analytical codes are not considered essential. The standard committee of Atomic Energy Society of Japan(AESJ) has drawn up the standard for using best estimate codes for safety analyses in 2008 after three-years of discussions reflecting domestic and international recent findings. (author)

  3. Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

    1985-04-01

    The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated

  4. Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

    Filbert, Wolfgang; Herold, Philipp

    2015-01-01

    The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.

  5. Solubility of radionuclides in a bentonite environment for provisional safety analyses for SGT-E2

    Berner, U.

    2014-08-01

    Within stage 2 of the sectoral plan for deep geological repositories for radioactive waste in Switzerland provisional safety analyses are carried out. In the case of the repository for spent fuel and vitrified high level waste considered, retention mechanisms include the concentration limits of safety relevant elements in the pore water of the buffer material (bentonite). The present work describes the solubility limits of the safety relevant elements Be, C_i_n_o_r_g, Cl, K, Ca, Co, Ni, Se, Sr, Zr, Nb, Mo, Tc, Pd, Ag, Sn, I, Cs, Sm, Eu, Ho, Pb, Po, Ra, Ac, Th, Pa, U, Np, Pu, Am and Cm in the pore water of bentonite after diffusive solution exchange with the host rock Opalinus Clay. The term solubility limit denotes the maximum amount of an element dissolving in the pore solution of the considered chemical reference system. Chemical equilibrium thermodynamics is the classical tool used for quantifying such considerations. For a given solid phase equilibrium thermodynamics predict the amount of substance dissolving in the solution and describe the speciation of the considered element in solution. The principles of chemical equilibrium will also be the primary work hypothesis in the present work. Solubility calculations were performed with the most recent version of GEMS/PSI (GEMS3.2 v.890) using the PSI/Nagra Chemical Thermodynamic Data Base 12/07, which is an update of the former Nagra/PSI Chemical Thermodynamic Data Base 01/01. The database was complemented with datasets from the ThermoChimie v. 7b for elements that were not considered in the mentioned update (Ag, Co, Sm, Ho, Pa, Be), with data from Iupac (Pb) and with data from the literature (Mo). Differing sources for thermodynamic data are noted. Reference values as well as lower and upper guideline values are evaluated. For many formation constants of solids and solutes uncertainties are known and allow conveying lower and upper guideline values. In many cases it is not clear whether the most stable solid is

  6. Safety

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  7. Initial Demonstration of the Real-Time Safety Monitoring Framework for the National Airspace System Using Flight Data

    Roychoudhury, Indranil; Daigle, Matthew; Goebel, Kai; Spirkovska, Lilly; Sankararaman, Shankar; Ossenfort, John; Kulkarni, Chetan; McDermott, William; Poll, Scott

    2016-01-01

    As new operational paradigms and additional aircraft are being introduced into the National Airspace System (NAS), maintaining safety in such a rapidly growing environment becomes more challenging. It is therefore desirable to have an automated framework to provide an overview of the current safety of the airspace at different levels of granularity, as well an understanding of how the state of the safety will evolve into the future given the anticipated flight plans, weather forecast, predicted health of assets in the airspace, and so on. Towards this end, as part of our earlier work, we formulated the Real-Time Safety Monitoring (RTSM) framework for monitoring and predicting the state of safety and to predict unsafe events. In our previous work, the RTSM framework was demonstrated in simulation on three different constructed scenarios. In this paper, we further develop the framework and demonstrate it on real flight data from multiple data sources. Specifically, the flight data is obtained through the Shadow Mode Assessment using Realistic Technologies for the National Airspace System (SMART-NAS) Testbed that serves as a central point of collection, integration, and access of information from these different data sources. By testing and evaluating using real-world scenarios, we may accelerate the acceptance of the RTSM framework towards deployment. In this paper we demonstrate the framework's capability to not only estimate the state of safety in the NAS, but predict the time and location of unsafe events such as a loss of separation between two aircraft, or an aircraft encountering convective weather. The experimental results highlight the capability of the approach, and the kind of information that can be provided to operators to improve their situational awareness in the context of safety.

  8. Cocaine Hydrolase Gene Transfer Demonstrates Cardiac Safety and Efficacy against Cocaine-Induced QT Prolongation in Mice

    Murthy, Vishakantha; Reyes, Santiago; Geng, Liyi; Gao, Yang; Brimijoin, Stephen

    2016-01-01

    Cocaine addiction is associated with devastating medical consequences, including cardiotoxicity and risk-conferring prolongation of the QT interval. Viral gene transfer of cocaine hydrolase engineered from butyrylcholinesterase offers therapeutic promise for treatment-seeking drug users. Although previous preclinical studies have demonstrated benefits of this strategy without signs of toxicity, the specific cardiac safety and efficacy of engineered butyrylcholinesterase viral delivery remains...

  9. Child Safety Seats on Commercial Airliners: A Demonstration of Cross-Price Elasticities

    Sanders, Shane; Weisman, Dennis L.; Li, Dong; Grimes, Paul, Ed.

    2008-01-01

    The cross-price elasticity concept can be difficult for microeconomics students to grasp. The authors provide a real-life application of cross-price elasticities in policymaking. After a debate that spanned more than a decade and included input from safety engineers, medical personnel, politicians, and economists, the Federal Aviation…

  10. Coupling of channel thermalhydraulics and fuel behaviour in ACR-1000 safety analyses

    Huang, F.L.; Lei, Q.M.; Zhu, W.; Bilanovic, Z.

    2008-01-01

    Channel thermalhydraulics and fuel thermal-mechanical behaviour are interlinked. This paper describes a channel thermalhydraulics and fuel behaviour coupling methodology that has been used in ACR-1000 safety analyses. The coupling is done for all 12 fuel bundles in a fuel channel using the channel thermalhydraulics code CATHENA MOD-3.5d/Rev 2 and the transient fuel behaviour code ELOCA 2.2. The coupling approach can be used for every fuel element or every group of fuel elements in the channel. Test cases are presented where a total of 108 fuel element models are set up to allow a full coupling between channel thermalhydraulics and detailed fuel analysis for a channel containing a string of 12 fuel bundles. An additional advantage of this coupling approach is that there is no need for a separate detailed fuel analysis because the coupling analysis, once done, provides detailed calculations for the fuel channel (fuel bundles, pressure tube, and calandria tube) as well as all the fuel elements (or element groups) in the channel. (author)

  11. Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

    2014-12-01

    In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

  12. Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

    NONE

    2014-12-15

    In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

  13. RECOMMENDED TRITIUM OXIDE DEPOSITION VELOCITY FOR USE IN SAVANNAH RIVER SITE SAFETY ANALYSES

    Lee, P.; Murphy, C.; Viner, B.; Hunter, C.; Jannik, T.

    2012-04-03

    The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95th percentile MOI exposure

  14. Demonstration of Risk Profiling for promoting safety in SME´s

    Jørgensen, Kirsten; Duijm, Nijs Jan; Troen, Hanne

    2011-01-01

    Purpose – The purpose of this paper is to identify and assess the risks and potential risks that may lead to accidents. It aims to look at how to improve risk assessment within SMEs for the benefit of all staff. Design/methodology/approach – The research included results from a Dutch project which...... identifies accident risks and safety barriers that are presented in a huge database and risk calculator. The method was first to develop a simple way of accessing this enormous amount of data, second, to develop a tool to observe risks and safety barriers in SMEs and to investigate the usefulness...... of the developed tools in real life, third, to collect data on risks and safety barriers in SMEs for two occupations by following 20 people for three days each and to create a risk profile for each occupations. Findings – The result is a simple way to go through all types of risks for accidents – a tool for risk...

  15. Environment, safety and health, management and organization compliance assessment, West Valley Demonstration Program, West Valley, New York

    1989-08-01

    An Environment, Safety and Health ''Tiger Team'' Assessment was conducted at the West Valley Demonstration Project. The Tiger Team was chartered to conduct an onsite, independent assessment of WVDP's environment, safety and health (ES ampersand H) programs to assure compliance with applicable Federal and State laws, regulations, and standards, and Department of Energy Orders. The objective is to provide to the Secretary of Energy the following information: current ES ampersand H compliance status of each facility; specific noncompliance items; ''root causes'' for noncompliance items; evaluation of the adequacy of ES ampersand H organization and resources (DOE and contractor) and needed modifications; and where warranted, recommendations for addressing identified problem areas

  16. Analysing supercritical water reactor's (SCWR's) special safety systems using probabilistic tools

    Ituen, I.; Novog, D.R.

    2011-01-01

    The next generation of reactors, termed Generation IV, has very attractive features -- its superior safety characteristics, high thermal efficiency, and fuel cycle sustainability. A key element of the Generation IV designs is the improvement in safety, which in turn requires improvements in safety system performance and reliability, as well as a reduction in initiating event frequencies. This study compares the response of the systems important to safety in the CANDU-Supercritical Water Reactor to those of the generic CANDU under a main steamline break accident and loss of forced circulation events -- to quantify the improvements in safety for the pre-conceptual CANDU SCWR design. Probabilistic safety analysis is the tool used in this study to test the behavior of the pre- conceptual design during these events. (author)

  17. Creativity Support to Improve Health-and-Safety in Manu-facturing Plants: Demonstrating Everyday Creativity

    Zachos, K.; Maiden, N.; Levis, S.

    2015-01-01

    This paper reports the development and deployment of digi-tal support for human creativity in a domain outside of the creative industries -- health-and-safety management in man-ufacturing plants. It reports applied research to extend a risk detection and resolution process at a world-class manufac-turing plant that produces tractors with creativity techniques and new digital support for the plant employees to use these techniques effectively as part of the risk detection and reso-lution proce...

  18. Utilisation of best estimate system codes and best estimate methods in safety analyses of VVER reactors in the Czech Republic

    Macek, Jiri; Kral, Pavel

    2010-01-01

    The content of the presentation was as follows: Conservative versus best estimate approach, Brief description and selection of methodology, Description of uncertainty methods, Examples of the BE methodology. It is concluded that where BE computer codes are used, uncertainty and sensitivity analyses should be included; if best estimate codes + uncertainty are used, the safety margins increase; and BE + BSA is the next step in licensing analyses. (P.A.)

  19. Demonstration of Emulator-Based Bayesian Calibration of Safety Analysis Codes: Theory and Formulation

    Joseph P. Yurko

    2015-01-01

    Full Text Available System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC sampling feasible. This work uses Gaussian Process (GP based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.

  20. Results of the safety analyses for the Greifswald and Stendal WWER nuclear power plants

    Milhem, J.L.

    1993-03-01

    Following a brief introduction of the design features of the three types of the WWER reactors, the paper deals with the main issues of the safety-related design and the most important recommendations which have been derived for upgrading measures. Furthermore some operational safety aspects of the VVER-1000 will be discussed in some detail

  1. Recommended safety procedures for the selection and use of demonstration-type gas discharge devices in schools

    1979-01-01

    A 1972 survey of 30 Ottawa secondary schools revealed a total of 347 actual or potential X-ray sources available in these schools. More than half of these sources were gas discharge tubes. Some gas discharge tubes, in particular the cold cathode type, can emit X-rays at significantly high levels. Unless such tubes are used carefully, and with regard for good radiation safety practices, they can result in exposures to students that are in excess of the maximum levels recommended by the International Commission on Radiological Protection. Several cases of the recommended dose being exceeded were found in the classes surveyed. This document has been prepared to assist science teachers and others using demonstration-type gas discharge devices to select and use such devices so as to present negligible risk to themselves and students. Useful information on safety procedures to be followed when performing demonstrations or experiments is included. (J.T.A.)

  2. The Tapioca Bomb: A Demonstration to Enhance Learning about Combustion and Chemical Safety

    Keeratichamroen, Wasana; Dechsri, Precharn; Panijpan, Bhinyo; Ruenwongsa, Pintip

    2010-01-01

    In any demonstration to students, producing light and sound usually ensures interest and can enhance understanding and retention of the concepts involved. A guided inquiry (Predict, Observe, Explain: POE) approach was used to involve the students actively in their learning about the explosive combustion of fine flour particles in air in the…

  3. Unique differences in applying safety analyses for a graphite moderated, channel reactor

    Moffitt, R.L.

    1993-06-01

    Unlike its predecessors, the N Reactor at the Hanford Site in Washington State was designed to produce electricity for civilian energy use as well as weapons-grade plutonium. This paper describes the major problems associated with applying safety analysis methodologies developed for commercial light water reactors (LWR) to a unique reactor like the N Reactor. The focus of the discussion is on non-applicable LWR safety standards and computer modeling/analytical variances of standards. The approaches used to resolve these problems to develop safety standards and limits for the N Reactor are described

  4. Difficulties in using Material Safety Data Sheets to analyse occupational exposures to contact allergens

    Friis, Ulrik F; Menné, Torkil; Flyvholm, Mari-Ann

    2015-01-01

    BACKGROUND: Information on the occurrence of contact allergens and irritants is crucial for the diagnosis of occupational contact dermatitis. Material Safety Data Sheets (MSDS) are important sources of information concerning exposures in the workplace. OBJECTIVE: From a medical viewpoint...

  5. Solubility of radionuclides in a concrete environment for provisional safety analyses for SGT-E2

    Berner, U.

    2014-08-15

    Within stage 2 of the sectoral plan for deep geological repositories for radioactive waste in Switzerland, safety analyses are carried out. In the case of the repository for long lived intermediate level waste (ILW) retention mechanisms include the concentration limits of safety relevant elements in the pore water of the engineered concrete system. The present work describes the evaluation of solubility limits for the safety relevant elements Be, C, Cl, K, Ca, Co, Ni, Se, Sr, Zr, Nb, Mo, Tc, Pd, Ag, Sn, I, Cs, Sm, Eu, Ho, Pb, Po, Ra, Ac, Th, Pa, U, Np, Pu, Am and Cm in the pore water of a concrete system corresponding to a degradation stage characterised by portlandite saturation and by the absence of (Na,K)OH solutes. The term solubility limit denotes the maximum amount of an element dissolving in the pore solution of the considered chemical reference system. For a given solid phase equilibrium, thermodynamics predicts the amount of substance dissolving in the solution and describes the speciation of the considered element in solution. The principles of chemical equilibrium will be the primary work hypothesis in the present work. Solubility calculations were performed with the most recent version of GEMS/PSI using the PSI/Nagra Chemical Thermodynamic Data Base 12/07. The database was complemented with other datasets for elements that were not considered in the mentioned update. Reference values solubilities as well as lower and upper guideline values are evaluated. For many formation constants of solids and solutes, uncertainties are known and allow conveying lower and upper guideline values. In many cases it is not clear whether the most stable solid is formed. In such cases the (kinetically driven) formation of alternative solid phases is included in the derivation of reference and guideline values. Corresponding justifications are given for the individual elements and are an important part of this work. A similar report for an almost identical chemical

  6. Solubility of radionuclides in a concrete environment for provisional safety analyses for SGT-E2

    Berner, U.

    2014-08-01

    Within stage 2 of the sectoral plan for deep geological repositories for radioactive waste in Switzerland, safety analyses are carried out. In the case of the repository for long lived intermediate level waste (ILW) retention mechanisms include the concentration limits of safety relevant elements in the pore water of the engineered concrete system. The present work describes the evaluation of solubility limits for the safety relevant elements Be, C, Cl, K, Ca, Co, Ni, Se, Sr, Zr, Nb, Mo, Tc, Pd, Ag, Sn, I, Cs, Sm, Eu, Ho, Pb, Po, Ra, Ac, Th, Pa, U, Np, Pu, Am and Cm in the pore water of a concrete system corresponding to a degradation stage characterised by portlandite saturation and by the absence of (Na,K)OH solutes. The term solubility limit denotes the maximum amount of an element dissolving in the pore solution of the considered chemical reference system. For a given solid phase equilibrium, thermodynamics predicts the amount of substance dissolving in the solution and describes the speciation of the considered element in solution. The principles of chemical equilibrium will be the primary work hypothesis in the present work. Solubility calculations were performed with the most recent version of GEMS/PSI using the PSI/Nagra Chemical Thermodynamic Data Base 12/07. The database was complemented with other datasets for elements that were not considered in the mentioned update. Reference values solubilities as well as lower and upper guideline values are evaluated. For many formation constants of solids and solutes, uncertainties are known and allow conveying lower and upper guideline values. In many cases it is not clear whether the most stable solid is formed. In such cases the (kinetically driven) formation of alternative solid phases is included in the derivation of reference and guideline values. Corresponding justifications are given for the individual elements and are an important part of this work. A similar report for an almost identical chemical

  7. Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

    Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

    2013-01-01

    Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

  8. Audiences, rationales and quantitative measure for demonstrations of nuclear safety and licensing by tests

    Lidsky, L.M.

    1990-01-01

    Nuclear power is one of several potential prime movers under consideration for central station production of electricity. As with any technology, the extent of its utilization depends on a complex set of interactions determined by its particular physical embodiments and the structure and temper of the society in which its use is considered. This paper focuses on the situation in the United States; its conclusions cannot easily be extrapolated to other nations. The interplay of indigenous resource base, political structure, and history is complex and must be analyzed case-by-case. I believe that the development of nuclear power plants with the ability to survive a definitive worst-case, 'absolute', test is a minimum requirement if nuclear power is to play a significant role in the future. The test protocols are somewhat dependent upon plant design, but include, at a minimum, simultaneous loss of coolant, control rod withdrawal, and the presence of a malicious operator. The test requirements are not determined by cost-benefit analysis nor by the imposition of mandated safety goals. They are substantially more stringent than would be required to meet even the most conservative commercial standards. Nonetheless, imposition of an absolute test is essential if the social and political prerequisites for the utilization of nuclear power are to be put in place. There are, of course, many other essential conditions, low cost being prime among them. The de facto imposition of an absolute test requirement would have several notable beneficial side effects: It would, for example, change the role of the NRC to one that has far greater public acceptance and it would lead to 'market force' standardization with attendant commercial ramifications

  9. An example demonstrating the conservatism of pipe calculations using KTA safety standard 3201.2

    Zeitner, W.

    1991-01-01

    The conservatism of the code calculation is demonstrated by using an example of a highly stressed pipe subject to internal pressure and a dynamic bending moment. For this reason the allowable code loadings are compared with the load carrying capacity, which is derived by realistic analysis (plastic strains) and experiment. The latter analysis is based on measured stress-strain curves of materials and stresses at which crack initiation occurs. The experiment shows that the pipe is capable of withstanding considerably higher loads than the code permits. The realistic analysis explains this discrepancy. (orig.)

  10. Program management plan for the conduct of a research, development, and demonstration program for improving the safety of nuclear powerplants

    1981-12-01

    Congress passed Public Law 96-567, Nuclear Safety Research, Development, and Demonstration Act of 1980, (hereafter referred to as the Act) to provide for an accelerated and coordinated program of light water reactor safety research, development, and demonstration to be carried out by the Department of Energy. In order to assure that this program would be compatible with the needs of Nuclear Regulatory Commission (NRC) and industry, the Department of Energy (DOE) initiated its response to Section 4 of the Act by conducting individual information gathering meetings with NRC and a wide cross section of the nuclear industry. The Department received recommendations on needs of what type of activities would and would not be appropriate for the Department to assist in satisfying these needs. Based on the evaluation of these inputs, it is concluded that the Department's ongoing Light Water Reactor (LWR) safety program is responsive to the Act. Specifically, the Department's ongoing program includes tasks in the areas of regulatory assessment, risk assessment, fission product source term, and emergency preparedness as well as providing technical assistance to the Institute of Nuclear Power Operations (INPO) to improve training of nuclear power personnel. These were among the very high priority efforts that were identified as necessary and appropriate for support by the Department

  11. A novel approach to delayed-start analyses for demonstrating disease-modifying effects in Alzheimer's disease.

    Hong Liu-Seifert

    Full Text Available One method for demonstrating disease modification is a delayed-start design, consisting of a placebo-controlled period followed by a delayed-start period wherein all patients receive active treatment. To address methodological issues in previous delayed-start approaches, we propose a new method that is robust across conditions of drug effect, discontinuation rates, and missing data mechanisms. We propose a modeling approach and test procedure to test the hypothesis of noninferiority, comparing the treatment difference at the end of the delayed-start period with that at the end of the placebo-controlled period. We conducted simulations to identify the optimal noninferiority testing procedure to ensure the method was robust across scenarios and assumptions, and to evaluate the appropriate modeling approach for analyzing the delayed-start period. We then applied this methodology to Phase 3 solanezumab clinical trial data for mild Alzheimer's disease patients. Simulation results showed a testing procedure using a proportional noninferiority margin was robust for detecting disease-modifying effects; conditions of high and moderate discontinuations; and with various missing data mechanisms. Using all data from all randomized patients in a single model over both the placebo-controlled and delayed-start study periods demonstrated good statistical performance. In analysis of solanezumab data using this methodology, the noninferiority criterion was met, indicating the treatment difference at the end of the placebo-controlled studies was preserved at the end of the delayed-start period within a pre-defined margin. The proposed noninferiority method for delayed-start analysis controls Type I error rate well and addresses many challenges posed by previous approaches. Delayed-start studies employing the proposed analysis approach could be used to provide evidence of a disease-modifying effect. This method has been communicated with FDA and has been

  12. Improving the safety of a body composition analyser based on the PGNAA method

    Miri-Hakimabad, Hashem; Izadi-Najafabadi, Reza; Vejdani-Noghreiyan, Alireza; Panjeh, Hamed [FUM Radiation Detection And Measurement Laboratory, Ferdowsi University of Mashhad (Iran, Islamic Republic of)

    2007-12-15

    The {sup 252}Cf radioisotope and {sup 241}Am-Be are intense neutron emitters that are readily encapsulated in compact, portable and sealed sources. Some features such as high flux of neutron emission and reliable neutron spectrum of these sources make them suitable for the prompt gamma neutron activation analysis (PGNAA) method. The PGNAA method can be used in medicine for neutron radiography and body chemical composition analysis. {sup 252}Cf and {sup 241}Am-Be sources generate not only neutrons but also are intense gamma emitters. Furthermore, the sample in medical treatments is a human body, so it may be exposed to the bombardments of these gamma-rays. Moreover, accumulations of these high-rate gamma-rays in the detector volume cause simultaneous pulses that can be piled up and distort the spectra in the region of interest (ROI). In order to remove these disadvantages in a practical way without being concerned about losing the thermal neutron flux, a gamma-ray filter made of Pb must be employed. The paper suggests a relatively safe body chemical composition analyser (BCCA) machine that uses a spherical Pb shield, enclosing the neutron source. Gamma-ray shielding effects and the optimum radius of the spherical Pb shield have been investigated, using the MCNP-4C code, and compared with the unfiltered case, the bare source. Finally, experimental results demonstrate that an optimised gamma-ray shield for the neutron source in a BCCA can reduce effectively the risk of exposure to the {sup 252}Cf and {sup 241}Am-Be sources.

  13. The Fort McMurray Demonstration Project in social marketing: health- and safety-related behaviour among oil sands workers.

    Guidotti, T L; Watson, L; Wheeler, M; Jhangri, G S

    1996-08-01

    This is the first round in a series of surveys conducted in Fort McMurray as part of the Fort McMurray Demonstration Project in social marketing. This component of the survey was intended to focus on the most prominent group of employed workers in the community and to compare their patterns of response with the community as a whole. Respondents to the survey were overwhelmingly male (96%), married (72.9%) and living in households of two to five persons (87.9%). They were predominantly aged 30-44 (55%) and graduates of high school (53.5%). Younger male workers (below age 30) were more likely to have a high school diploma (78.3%) or some additional technical or vocational training (21.7% compared to 12.5% overall) and to be unmarried or separated. Attitudes toward safety-related behaviours were stronger than for respondents from the community as a whole. Approximately 70-100% of all age groups and both sexes showed strong agreement with attitudes involving child car seats and the unacceptability of drinking and driving. These attitudes include strong advocacy of vigorous enforcement of occupational health and safety standards. However, they showed a variability similar to the community as a whole in behaviour at home compared to work, generally reporting more consistent use of personal protection on the job than in their own homes, particularly hearing protection. Even so, they were much less likely to perform stretching and warm-up exercises prior to exertion than community residents in general. The potential may exist to transfer the technology and attitudes from workplace health and safety to community safety. One possible strategy to accomplish this is to involve workers in this industry directly in community initiatives. This strategy may be generalizable to any community in which there are major employers who place a heavy emphasis on risk control and occupational health and safety.

  14. [Patient safety and errors in medicine: development, prevention and analyses of incidents].

    Rall, M; Manser, T; Guggenberger, H; Gaba, D M; Unertl, K

    2001-06-01

    "Patient safety" and "errors in medicine" are issues gaining more and more prominence in the eyes of the public. According to newer studies, errors in medicine are among the ten major causes of death in association with the whole area of health care. A new era has begun incorporating attention to a "systems" approach to deal with errors and their causes in the health system. In other high-risk domains with a high demand for safety (such as the nuclear power industry and aviation) many strategies to enhance safety have been established. It is time to study these strategies, to adapt them if necessary and apply them to the field of medicine. These strategies include: to teach people how errors evolve in complex working domains and how types of errors are classified; the introduction of critical incident reporting systems that are free of negative consequences for the reporters; the promotion of continuous medical education; and the development of generic problem-solving skills incorporating the extensive use of realistic simulators wherever possible. Interestingly, the field of anesthesiology--within which realistic simulators were developed--is referred to as a model for the new patient safety movement. Despite this proud track record in recent times though, there is still much to be done even in the field of anesthesiology. Overall though, the most important strategy towards a long-term improvement in patient safety will be a change of "culture" throughout the entire health care system. The "culture of blame" focused on individuals should be replaced by a "safety culture", that sees errors and critical incidents as a problem of the whole organization. The acceptance of human fallability and an open-minded non-punitive analysis of errors in the sense of a "preventive and proactive safety culture" should lead to solutions at the systemic level. This change in culture can only be achieved with a strong commitment from the highest levels of an organization. Patient

  15. An integrated software system for core design and safety analyses: Cascade-3D

    Wan De Velde, A.; Finnemann, H.; Hahn, T.; Merk, S.

    1999-01-01

    The new Siemens program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of the most advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management. (authors)

  16. Safety demonstration tests of postulated solvent fire accidents in extraction process of a fuel reprocessing plant, (2)

    Tukamoto, Michio; Takada, Junichi; Koike, Tadao; Nishio, Gunji; Uno, Seiichiro; Kamoshida, Atsusi; Watanabe, Hironori; Hashimoto, Kazuichiro; Kitani, Susumu.

    1992-03-01

    Demonstration tests of hypothetical solvent fire in an extraction process of the reprocessing plant were carried out from 1984 to 1985 in JAERI, focusing on the confinement of radioactive materials during the fire by a large-scale fire facility (FFF) to evaluate the safety of air-ventilation system in the plant. Fire data from the demonstration test were obtained by focusing on fire behavior at cells and ducts in the ventilation system, smoke generation during the fire, transport and deposition of smoke containing simulated radioactive species in the ventilation system, confinement of radioactive materials, and integrity of HEPA filters by using the FFF simulating an air-ventilation system of the reference reprocessing plant in Japan. The present report is published in a series of the report Phase I (JAERI-M 91-145) of the demonstration test. Test results in the report will be used for the verification of a computer code FACE to evaluate the safety of postulated fire accidents in the reprocessing plant. (author)

  17. A Prototype Lip Balm: Summary of Three Dermatological Studies Demonstrating Safety and Acceptability for Sensitive Skin.

    Nisbet, Stephanie

    Data were generated from three studies to assess the tolerability and acceptability of a prototype cosmetic lip balm. Dermatological assessments of topical compatibility (primary and cumulative irritability and sensitization), photoirritant and topical photosensitizer potential, and acceptability for safe use of a prototype cosmetic lip balm on sensitive skin are summarized. In Study 1, the product was applied to the volunteers' backs under a semiocclusive patch followed by patch removal/reapplication over 6 weeks to assess the irritant and allergic potential of the product. Dermatological assessments were performed at the beginning and end of the study or when there was evidence of positivity or adverse event. Study 2 was conducted by applying the product to the volunteers' backs under a semiocclusive patch, followed by patch removal/reapplication and irradiation of the test area with ultraviolet A (UVA) radiation at various intervals over 5 weeks. Dermatological assessments were performed to assess the product's role in the induction of photoirritancy and photosensitization. Clinical and subjective assessments for acceptability were obtained during Study 3 in volunteers with a diagnosis of sensitive skin and those who used the product as per instructions for use during the study period. The data generated from the three studies demonstrated no evidence of primary or cumulative dermal irritation or of dermal sensitization. In addition, no photoirritation potential or photosensitization potential was observed. As assessed by dermatologic monitoring and subject diary entries, the prototype lip balm did not cause irritation or sensitization reactions when used for 28 days in volunteers with a diagnosis of sensitive skin. Based on these findings, the prototype lip balm can be considered suitable for use for people with sensitive skin.

  18. Sensitivity and Uncertainty Analyses Applied to Neutronics Calculations for Safety Assessment at IRSN

    Ivanov, Evgeny; Ivanova, Tatiana; Pignet, Sophie

    2013-01-01

    Objective of the presentation: • Present IRSN vision relevant to validation of stand-alone neutronics codes on support of the fuel cycle and reactor safety assessment for fast neutron reactors. • Provide work status, future developments and needs for R&D working program on validation methodology for neutronics of fast systems

  19. Elemental analyses of goundwater: demonstrated advantage of low-flow sampling and trace-metal clean techniques over standard techniques

    Creasey, C. L.; Flegal, A. R.

    'introduction accidentelle de contaminants au cours de l'échantillonnage, du stockage et de l'analyse. Lorsque ces techniques sont appliquées, les concentrations résultantes en éléments en traces sont nettement plus faibles que les résultats obtenus par les techniques d'échantillonnage classique. Dans une comparaison de données concernant des puits contaminés et des puits de contrôle d'un site de Californie (États-Unis), les concentrations en éléments en traces de cette étude ont été de 2 à 1000 fois plus faibles que celles déterminées par les techniques conventionnelles utilisées pour l'échantillonnage des mêmes puits cinq mois auparavant et un mois après ces prélèvements. En particulier, les concentrations en cadmium et en chrome obtenues par les techniques classiques de prélèvements dépassent les teneurs maximales admises en Californie, alors que les concentrations obtenues pour ces deux éléments dans cette étude sont nettement au-dessous de ces teneurs maximales. Par conséquent, le recours à des techniques à faible débit et sans traces de métal peut faire apparaître que la publication de contamination d'eaux souterraines par des éléments en traces était erronée. Resumen El uso combinado del purgado y muestreo a bajo caudal con las técnicas limpias de metales traza proporcionan medidas de la concentración de elementos traza en las aguas subterráneas que son más representativas que las obtenidas con técnicas tradicionales. El purgado y muestreo a bajo caudal proporciona muestras de agua prácticamente inalteradas, representativas de las condiciones en el terreno. Las técnicas limpias de metales traza limitan la no deseada introducción de contaminantes durante el muestreo, almacenamiento y análisis. Las concentraciones de elementos traza resultantes suelen ser bastante menores que las obtenidas por técnicas tradicionales. En una comparación entre los datos procedentes de pozos en California, las concentraciones obtenidas con el nuevo m

  20. Archaeometrical analyses demonstrates that humans excavated clay from mardels on the Luxembourger Gutland plateau to produce ceramics.

    Vanmourik, Jan; Braekmans, Dennis

    2017-04-01

    sediments must have been used, most probably for the production of ceramics. If we can find relicts of ceramics in the vicinity of mardels we can compare the composition of these ceramics with mardel clay. We could collect finds of Roman tile-works on the Lias marls (Kalefeld) and of Roman pottery on the Keuper marls (Biischtert) in the vicinity of mardels. Provenance analysis (XRF) demonstrated the similarity of chemical composition of mardel clay and ceramics. This indicates that the mardels on the Gutland plateau developed initially as natural depressions (sediment traps), on Strassen marls related to soil subsidence, caused by joints in the underlying Luxembourger sandstone, on Keuper marls to subsidence after soil subsurface dissolving of gypsum veins. The colluvial clay was used by the Romans for the production of ceramics. Due to the excavation reached the actual seizes. Colluvial clay accumulation restarted in the abandoned quarries. • Slotboom, R.T. (1963) Comparative geomorphological and palynological investigation of the pingos (Viviers) in the Haute Fagnes (Belgium) and the Mardellen in the Gutland (Luxembourg). Zeitschrift für Geomorphologie 7: 193-231. • Schmalen, C. (2002) Einige Mardellen Luxemburgs auf den Keuper-und Liasschichten des Forstamtbezirks Zentrum. Diplomarbeit in Studiengang Umweltplannung an der Fachhochschule Trier, Standort Birkenfeld, 2002. • Etienne, D., Ruffaldi, P., Goepp, S., Ritz, F., Georges-Leroy, M., Pollier, B., Dambrine, E. (2011) The origin of closed depressions in Northeastern France: A new assessment. Geomorphology 126: 121-131. • Slotboom, R.T., van Mourik, J.M. (2015) Pollen records of mardel deposits; the effects of climatic oscillations and land management on soil erosion in Gutland, Luxembourg. Catena 132 (2015) 72-88.

  1. Assessment of S(α, β) libraries for criticality safety evaluations of wet storage pools by refined trend analyses

    Kolbe, E.; Vasiliev, A.; Ferroukhi, H.

    2009-01-01

    In a recent criticality safety evaluation (CSE) of a commercial wet storage pool applying MCNPX-2.5.0 in combination with the ENDF/B-VII.0 and JEFF-3.1 continuous energy cross section libraries, the maximum permissible initial fuel-enrichment limit for water reflected configurations was found to be dependant upon the applied neutron cross section library. More detailed investigations indicated that the difference is mainly caused by different sub-libraries for thermal neutron scattering based on parameterizations of the S(α, β) scattering matrix. Hence an analysis of trends was done with respect to the low energy neutron flux in order to assess the S(α, β) data sets. First, when performing the trend analysis based on the full set of 149 benchmarks that were employed for the validation, significant trends could not be found. But by analyzing a selected subset of benchmarks clear trends with respect to the low energy neutron flux could be detected. The results presented in this paper demonstrate the sensitivity of specific configurations to the parameterizations of the S(α, β) scattering matrix and thus may help to improve CSE of wet storage pools. Finally, in addition to the low energy neutron flux, we also refined the trend analyses with respect to other key (spectrum-related) parameters by performing them with various selected subsets of the full suite of 149 benchmarks. The corresponding outcome using MCNPX 2.5.0 in combination with the ENDF/B-VII.0, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, and JENDL-3.3 neutron cross section libraries are presented and discussed. (authors)

  2. Development of advanced methods and related software for human reliability evaluation within probabilistic safety analyses

    Kosmowski, K.T.; Mertens, J.; Degen, G.; Reer, B.

    1994-06-01

    Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Analysis (PSA). The first part of this report consists of an overview of types of human behaviour and human error including the effect of significant performance shaping factors on human reliability. Particularly with regard to safety assessments for nuclear power plants a lot of HRA methods have been developed. The most important of these methods are presented and discussed in the report, together with techniques for incorporating HRA into PSA and with models of operator cognitive behaviour. Based on existing HRA methods the concept of a software system is described. For the development of this system the utilization of modern programming tools is proposed; the essential goal is the effective application of HRA methods. A possible integration of computeraided HRA within PSA is discussed. The features of Expert System Technology and examples of applications (PSA, HRA) are presented in four appendices. (orig.) [de

  3. Probabilistic safety analyses. Status and further development of methods and models, applications

    Berg, H.P.; Schott, H.

    1992-12-01

    The report describes the topics of the deterministic and probabilistic approach. The PSA is used in order to investigate event sequences beyond design limits; in particular the expected frequency of core melting is important. The basis of PSA is described including its limits. Moreover, the current state of the art of science and technology in the field of PSA including the so-called 'living PSA' are explained. Some measures which result in order to improve the safety of a nuclear power plant from the German Risk-Study are shown. An overview is given on the status of PSA in periodic safety reviews in German nuclear power plants. Moreover, the main topics of running investigations are presented. (orig.) [de

  4. Chapter 2: Development of instrumentation for safety analyses in fuel reprocessing and treatment plants

    Anon.

    1985-01-01

    Development and provision of methods allowing for safety-related statements on non-appropriate operation of intermediate storage, reprocessing and waste conditioning on the basis of probabilities. By applying the methods and models to the courses of events considered, activity releases at the chimney and their probable frequency were determined. For accidents known to be radiologically relevant, expected values for exposure were computed by means of complex distribution and exposure models. (DG) [de

  5. Recommended Tritium Oxide Deposition Velocity For Use In Savannah River Site Safety Analyses

    Lee, P. L.; Murphy, C. E.; Viner, B. J.; Hunter, C. H.

    2012-07-31

    This report documents the results of examining the deposition velocity of water to forests, the residence time of HTO in forests, and the relation between deposition velocity and residence time with specific consideration given to the topography and experimental work performed at SRS. A simple mechanistic model is used to obtain plausible deposition velocity and residence time values where experimental data are not available and recommendations are made for practical application in a safety analysis model.

  6. Sensitivity and uncertainty analyses applied to criticality safety validation, methods development. Volume 1

    Broadhead, B.L.; Hopper, C.M.; Childs, R.L.; Parks, C.V.

    1999-01-01

    This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the available S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently used by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The S/U methods that are presented in this volume are designed to provide a formal means of establishing the range (or area) of applicability for criticality safety data validation studies. The development of parameters that are analogous to the standard trending parameters forms the key to the technique. These parameters are the D parameters, which represent the differences by group of sensitivity profiles, and the ck parameters, which are the correlation coefficients for the calculational uncertainties between systems; each set of parameters gives information relative to the similarity between pairs of selected systems, e.g., a critical experiment and a specific real-world system (the application)

  7. An approach of sensitivity and uncertainty analyses methods installation in a safety calculation

    Pepin, G.; Sallaberry, C.

    2003-01-01

    Simulation of the migration in deep geological formations leads to solve convection-diffusion equations in porous media, associated with the computation of hydrogeologic flow. Different time-scales (simulation during 1 million years), scales of space, contrasts of properties in the calculation domain, are taken into account. This document deals more particularly with uncertainties on the input data of the model. These uncertainties are taken into account in total analysis with the use of uncertainty and sensitivity analysis. ANDRA (French national agency for the management of radioactive wastes) carries out studies on the treatment of input data uncertainties and their propagation in the models of safety, in order to be able to quantify the influence of input data uncertainties of the models on the various indicators of safety selected. The step taken by ANDRA consists initially of 2 studies undertaken in parallel: - the first consists of an international review of the choices retained by ANDRA foreign counterparts to carry out their uncertainty and sensitivity analysis, - the second relates to a review of the various methods being able to be used in sensitivity and uncertainty analysis in the context of ANDRA's safety calculations. Then, these studies are supplemented by a comparison of the principal methods on a test case which gathers all the specific constraints (physical, numerical and data-processing) of the problem studied by ANDRA

  8. Safety demonstration tests on thermal decomposition of nitrated solvent with nitric acid in nuclear fuel reprocessing plants. Contract research

    Tsukamoto, Michio; Takada, Junichi; Koike, Tadao; Watanabe, Koji; Uchiyama, Gunzou; Nishio, Gunji; Murata, Mikio

    2001-03-01

    The demonstration tests were conducted to investigate the safety of the ventilation system and integrity of the HEPA filters under the design basis accident (DBA) of the evaporator in the reprocessing plants. The tests were carried out by heating organic solvent (TBP/n- dodecane) mixed with nitric acid in a sealed vessel. It was possible to cause an explosive decomposition of TBP-complex formed by nitration of the solvent with nitric acid. The following was obtained by the analysis of the experimental results of the tests. From derivation by the experimental method, data on the maximum mass release rate and the maximum energy release rate in the explosion, as the solvent of 1 [kg] spouted out by the thermal decomposition, were obtained. They were 0.59 [kg/s] and 3240.3 [kJ/kg·s] respectively. The influence given on the cell ventilation system by this explosion was small and it was demonstrated that the safety of the HEPA filters could be secured. (author)

  9. Safety demonstration tests on pressure rise in ventilation system and blower integrity of a fuel-reprocessing plant

    Takada, Junichi; Suzuki, Motoe; Tsukamoto, Michio; Koike, Tadao; Nishio, Gunji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-12-01

    In JAERI, the demonstration test was carried out as a part of safety researches of the fuel-reprocessing plant using a large-scale facility consist of cells, ducts, dumpers, HEPA filters and a blower, when an explosive burning due to a rapid reaction of thermal decomposition for solvent/nitric acid occurs in a cell of the reprocessing plant. In the demonstration test, pressure response propagating through the facility was measured under a blowing of air from a pressurized tank into the cell in the facility to elucidate an influence of pressure rise in the ventilation system. Consequently, effective pressure decrease in the facility was given by a configuration of cells and ducts in the facility. In the test, transient responses of HEPA filters and the blower by the blowing of air were also measured to confirm the integrity. So that, it is confirmed that HEPA filters and the blower under pressure loading were sufficient to maintain the integrity. The content described in this report will contribute to safety assessment of the ventilation system in the event of explosive burning in the reprocessing plant. (author)

  10. Calculational framework for safety analyses of non-reactor nuclear facilities

    Coleman, J.R.

    1994-01-01

    A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks

  11. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

    1977-01-01

    The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

  12. Spacecraft Fire Safety Demonstration

    Urban, David L.; Ruff, Gary A.

    2016-01-01

    A presentation of the Saffire Experiment goals and scientific objectives for the Joint CSA/ESA/JAXA/NASA Increments 47 and 48 Science Symposium. The purpose of the presentation is to inform the ISS Cadre and the other investigators of the Saffire goals and objectives to enable them to best support a successful Saffire outcome.

  13. Risk and safety analyses for disposal of alpha-contaminated waste in INEL

    Smith, T.

    1982-01-01

    The author first discusses the context, objectives, and scope of the risk analysis. Then he gives some background on the waste and how its managed, including the alternatives for long-term management. These are followed by risk evaluation approach, results, and 7 conclusions and problems. One of his conclusions is that a 100 nCi/g limit would provide adequate safety margins. Raising the limit to 100 nCi/g would allow about 20% of the stored waste to be diverted to near-surface disposal. He added that analyzing waste packages at 10 nCi/g is not now practical. 21 figures

  14. Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

  15. LOCA, LOFA and LOVA analyses pertaining to NET/ITER safety design guidance

    Ebert, E.; Raeder, J.

    1991-01-01

    The analyses presented pertain to loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA). These types of accidents may jeopardise components and plasma vessel integrity and cause radioactivity mobilisation. The analyses reviewed have been performed under the assumption that the plasma facing components are protected by a carbon based armour. Accidental temperatures and pressure transients are quantified, the possibility of reaction products combustion is investigated and worst case accidental public doses are assessed. On this basis, design recommendations are given and design features such as low plasma facing components armour temperatures (on almost the entire surface) and inert gas adjacent to the vacuum vessel have been implemented. (orig.)

  16. [Supervised exercise training in patients with pulmonary arterial hypertension - analyses of the effectiveness and safety].

    Saxer, S; Rhyner, M; Treder, U; Speich, R; van Gestel, A J R

    2012-02-01

    Both in today's scientific research and in clinical practice, there exists a need to address the uncertainty concerning the effectiveness and safety of cardiopulmonary exercise training (CPET) in patients with pulmonary arterial hypertension (PAH). It is commonly believed that CPET may be dangerous for patients with PAH, because increasing pressure on the pulmonary arteries may worsen right-sided heart failure. Recently, the first clinical trials on exercise training in patients with pulmonary hypertension reported promising results. Extension of the walking distance at the 6-minute walk test improved quality of life, endurance capacity and a reduction in symptoms were observed after CPET. Furthermore, CPET was well tolerated by the patients in five clinical trials. In conclusion, it may be postulated that CPET is an effective therapy in patients with PAH and was tendentially well tolerated by the patients.

  17. Population analyses of efficacy and safety of ABT-594 in subjects with diabetic peripheral neuropathic pain.

    Dutta, Sandeep; Hosmane, Balakrishna S; Awni, Walid M

    2012-06-01

    ABT-594, a neuronal nicotinic acetylcholine receptor ligand, is 30- to 100-fold more potent than morphine in animal models of nociceptive and neuropathic pain. Efficacy and safety of ABT-594 in subjects with painful diabetic polyneuropathy was evaluated in a phase 2 study. The objective of this work was to use a nonlinear mixed effects model-based approach for characterizing the relationship between dose and response (efficacy and safety) of ABT-594. Subjects (N = 266) were randomized into four groups in a double-blind, placebo-controlled, 7-week study to receive twice daily regimens of placebo or 150, 225, and 300 μg of ABT-594. The primary efficacy variable, pain score (11-point Likert scale), was assessed on five occasions. The probability of change from baseline pain score of ≥1, ≥2, and ≥3 was modeled using cumulative logistic regression with dose and days of treatment as explanatory variables. The incidence of five most frequently occurring adverse events (AEs) was modeled using linear logistic regression. ABT-594 ED(50) values (improvement in 50% of subjects) for improvement in pain scores of ≥1, ≥2, and ≥3 were 50, 215, and 340 μg, respectively, for the average number of days (33) on treatment. The rank order of ED(50) values for AEs was nausea, vomiting, dizziness, headache, and abnormal dreams; nicotine users were less sensitive to AEs. Population pharmacodynamic models developed to characterize the improvement in pain score and incidence of adverse events indicate an approximately twofold separation between the ED(50) values for efficacy and AEs.

  18. Evaluation of geological documents available for provisional safety analyses of potential sites for nuclear waste repositories - Are additional geological investigations needed?

    2010-10-01

    . In order to set priorities based on safety, it is necessary to use the characteristic dose intervals determined with the help of dose calculations, an evaluation of engineering feasibility and the results of a qualitative assessment. The state of knowledge is evaluated using test calculations, which are used to derive dose curves for a wide spectrum of cases for the different repository types. Besides the expected evolution (reference case), the calculations also cover existing uncertainties in the relevant processes and parameters. This allows the characteristic dose intervals required by the Sectoral Plan to be determined for the different siting regions for each repository type; these dose intervals are evaluated with respect to their unambiguousness concerning two aspects required by the Sectoral Plan: safety-related suitability and safety-related equivalence of the siting regions. The state of knowledge is considered to be sufficient if clear, unambiguous statements can be made despite the fact that the parameter ranges are selected generously to allow for existing uncertainties and if these statements will not change if the uncertainties and associated parameter ranges are reduced through future investigations. Whether the information used to assess engineering feasibility is sufficient is also considered, along with an evaluation of the significance of uncertainties in the relevant processes and parameters. For all siting regions, concrete statements can be made regarding suitability and equivalence of sites from the point of view of safety despite the deliberate selection of wide parameter ranges to take into account uncertainties; engineering feasibility has also been demonstrated. This means that no additional investigations will be necessary for the provisional safety analyses to be performed in Stage 2 of the Sectoral Plan. The report also describes the work already initiated and planned by Nagra for Stage 2 of the Sectoral Plan. This work contributes to

  19. The importance of probabilistic evaluations in connection with risk analyses according to technical safety laws

    Mathiak, E.

    1984-01-01

    The nuclear energy sector exemplifies the essential importance to be attached to the practical application of probabilistic evaluations (e.g. probabilistic reliability analyses) in connection with the legal risk assessment of technical systems and installations. The study is making use of a triad risk analysis and tries to reconcile the natural science and legal points of view. Without changing the definitions of 'risk' and 'hazard' in the legal sense of their meaning the publication discusses their reconcilation with the laws of natural science, their interpretation and application in view of the latter. (HSCH) [de

  20. DNA Analyses in Food Safety and Quality: Current Status and Expectations

    Marchelli, Rosangela; Tedeschi, Tullia; Tonelli, Alessandro

    Food safety and quality are very important issues receiving a lot of attention in most countries by producers, consumers and regulatory and control authorities. In particular, DNA analysis in food is becoming popular not only in relation to genetically modified products (GMOs), in which DNA modification is the "clue" of the novelty, but also in other fields like microbiology and pathogen detection, which require long times for the cultivation and specially in cases in which the microorganisms are not cultivable like some viruses, as well as for authenticity and allergen detection. A new topic concerning "nutrigenetics and nutrigenomics" has also been mentioned, very important but still in its infancy, which could lead in the future to a personalized diet. In this chapter we have described the main areas of food research and fields of application where DNA analysis is being performed and the relative methods of detection, which are generally based on PCR. The possibility/opportunity to detect DNA without previous amplification (PCR-free) will be discussed. We have examined the following areas: (1) genetically modified foods (GMOs); (2) food allergens; (3) microbiological contaminations; (4) food authenticity; (5) nutrigenetics/nutrigenomics.

  1. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  2. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    Heard, F.J.

    1997-01-01

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

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

    1983-01-01

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

  4. Reliability analyses of safety systems for WWER-440 nuclear power plants

    Dusek, J.; Hojny, V.

    1985-01-01

    The UJV in Rez near Prague studied the reliability of the system of emergency core cooling and of the system for suppressing pressure in the sealed area of the nuclear power plant in the occurrence of a loss-of-coolant accident. The reliability of the systems was evaluated by failure tree analysis. Simulation and analytical calculation programs were developed and used for the reliability analysis. The results are briefly presented of the reliability analyses of the passive system for the immediate short-term flooding of the reactor core, of the active low-pressure system of emergency core cooling, the spray system, the bubble-vacuum system and the system of emergency supply of the steam generators. (E.S.)

  5. Development of SAGE, A computer code for safety assessment analyses for Korean Low-Level Radioactive Waste Disposal

    Zhou, W.; Kozak, Matthew W.; Park, Joowan; Kim, Changlak; Kang, Chulhyung

    2002-01-01

    This paper describes a computer code, called SAGE (Safety Assessment Groundwater Evaluation) to be used for evaluation of the concept for low-level waste disposal in the Republic of Korea (ROK). The conceptual model in the code is focused on releases from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. Doses can be calculated for several biosphere systems including drinking contaminated groundwater, and subsequent contamination of foods, rivers, lakes, or the ocean by that groundwater. The flexibility of the code will permit both generic analyses in support of design and site development activities, and straightforward modification to permit site-specific and design-specific safety assessments of a real facility as progress is made toward implementation of a disposal site. In addition, the code has been written to easily interface with more detailed codes for specific parts of the safety assessment. In this way, the code's capabilities can be significantly expanded as needed. The code has the capability to treat input parameters either deterministic ally or probabilistic ally. Parameter input is achieved through a user-friendly Graphical User Interface.

  6. Industrial Fuel Gas Demonstration Plant Program. Conceptual design and evaluation of commercial plant. Volume III. Economic analyses (Deliverable Nos. 15 and 16)

    None

    1978-01-01

    This report presents the results of Task I of Phase I in the form of a Conceptual Design and Evaluation of Commercial Plant report. The report is presented in four volumes as follows: I - Executive Summary, II - Commercial Plant Design, III - Economic Analyses, IV - Demonstration Plant Recommendations. Volume III presents the economic analyses for the commercial plant and the supporting data. General cost and financing factors used in the analyses are tabulated. Three financing modes are considered. The product gas cost calculation procedure is identified and appendices present computer inputs and sample computer outputs for the MLGW, Utility, and Industry Base Cases. The results of the base case cost analyses for plant fenceline gas costs are as follows: Municipal Utility, (e.g. MLGW), $3.76/MM Btu; Investor Owned Utility, (25% equity), $4.48/MM Btu; and Investor Case, (100% equity), $5.21/MM Btu. The results of 47 IFG product cost sensitivity cases involving a dozen sensitivity variables are presented. Plant half size, coal cost, plant investment, and return on equity (industrial) are the most important sensitivity variables. Volume III also presents a summary discussion of the socioeconomic impact of the plant and a discussion of possible commercial incentives for development of IFG plants.

  7. A review of significant events analysed in general practice: implications for the quality and safety of patient care

    Bradley Nick

    2009-09-01

    Full Text Available Abstract Background Significant event analysis (SEA is promoted as a team-based approach to enhancing patient safety through reflective learning. Evidence of SEA participation is required for appraisal and contractual purposes in UK general practice. A voluntary educational model in the west of Scotland enables general practitioners (GPs and doctors-in-training to submit SEA reports for feedback from trained peers. We reviewed reports to identify the range of safety issues analysed, learning needs raised and actions taken by GP teams. Method Content analysis of SEA reports submitted in an 18 month period between 2005 and 2007. Results 191 SEA reports were reviewed. 48 described patient harm (25.1%. A further 109 reports (57.1% outlined circumstances that had the potential to cause patient harm. Individual 'error' was cited as the most common reason for event occurrence (32.5%. Learning opportunities were identified in 182 reports (95.3% but were often non-specific professional issues not shared with the wider practice team. 154 SEA reports (80.1% described actions taken to improve practice systems or professional behaviour. However, non-medical staff were less likely to be involved in the changes resulting from event analyses describing patient harm (p Conclusion The study provides some evidence of the potential of SEA to improve healthcare quality and safety. If applied rigorously, GP teams and doctors in training can use the technique to investigate and learn from a wide variety of quality issues including those resulting in patient harm. This leads to reported change but it is unclear if such improvement is sustained.

  8. A tool for safety officers when analysing the basic causes of simple accidents

    Jørgensen, Kirsten

    Most accidents that happen in enterprises are simple and seldom have serious invalidating consequences. Very often these kinds of accident are not investigated and if they are, then the investigation is very brief, with comments such as that it was the victim’s own fault or just an unlucky...... for some years with interesting results. Both the difficulties and the benefits will be presented, together with examples of the use of the tool. The main purpose of the tool is to demonstrate how management and workers can get a much better understanding of why accidents happen, even those accidents...... that seem to be unavoidable, and that simple accidents never are simple, but always have root causes on which preventive action can be focused....

  9. Biosphere analyses for the safety assessment SR-Site - synthesis and summary of results

    Saetre, Peter

    2010-12-01

    ice free period between two glaciations. The radionuclide model used in SR-Site has been improved in several important ways since previous safety assessments conducted by SKB. For example, the aquatic and terrestrial ecosystems are handled in the same model, which gives a continuous transition from the sea stage to the lake and terrestrial stages. Transport and accumulation in till (lower regolith) is represented in the model. The uptake by plants is included in the mass-balance, and it is related to biomass growth. Moreover, parameter values including hydrological flows, sedimentation and resuspension rates, biomass growth rates, gas exchange rates, as well as element specific distribution coefficients and concentration rations, were as far as possible based on site data. One endpoint from the simulations with the radionuclide model was the landscape dose conversion factors (LDFs). The LDF represents the mean annual effective dose over lifetime for an individual living in the most contaminated area, assuming a constant unit release rate (1Bq/y). In the safety assessment, the maximum LDF for each nuclide have been selected from the biosphere object at the time yielding the highest unit release dose, and consequently LDFs from different nuclides does not necessarily match the same group of exposed individuals with respect to point in time or location in the landscape. In the SR-Site main report, the resulting dose is presented when the maximum LDF is multiplied by a release. The potential effect of a radionuclide release on non-human biota in Forsmark is also assessed

  10. Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

    2009-11-15

    deterioration over the assessment period. The basic approach for prescribing such margins is to consider whether the design assessed in SR-Can Main report was sufficient to result in safety. In case this design would imply too strict requirements, and in cases the SR-Can design was judged inadequate or not sufficiently analysed in the SR-Can report, some additional analyses have been undertaken to provide a better basis for setting the design premises. The resulting design premises constitute design constraints, which, if all fulfilled, form a good basis for demonstrating repository safety, according to the analyses in SR-Can and subsequent analyses. Some of the design premises may be modified in future stages of SKB's programme, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. Furthermore, a different balance between design requirements may result in the same level of safety. This report presents one technically reasonable balance, whereas future development and evaluations may result in other balances being deemed as more optimal. It should also be noted that in developing the reference design, the production reports should give credible evidence that the final product after construction and quality control fulfils the specifications of the reference design. To cover uncertainties in production and quality control that may be difficult to quantify in detail at the present design stage, the developer of the reference design need usually consider a margin to the conditions that would verify the design premises, but whether there is a need for such margins lies outside the scope of the current document. The term 'withstand' is used in this document in descriptions of load cases on repository components. The statement that a component withstands a particular load means that it upholds its related safety function when exposed to the load in question. For example, if the

  11. Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

    2009-11-01

    deterioration over the assessment period. The basic approach for prescribing such margins is to consider whether the design assessed in SR-Can Main report was sufficient to result in safety. In case this design would imply too strict requirements, and in cases the SR-Can design was judged inadequate or not sufficiently analysed in the SR-Can report, some additional analyses have been undertaken to provide a better basis for setting the design premises. The resulting design premises constitute design constraints, which, if all fulfilled, form a good basis for demonstrating repository safety, according to the analyses in SR-Can and subsequent analyses. Some of the design premises may be modified in future stages of SKB's programme, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. Furthermore, a different balance between design requirements may result in the same level of safety. This report presents one technically reasonable balance, whereas future development and evaluations may result in other balances being deemed as more optimal. It should also be noted that in developing the reference design, the production reports should give credible evidence that the final product after construction and quality control fulfils the specifications of the reference design. To cover uncertainties in production and quality control that may be difficult to quantify in detail at the present design stage, the developer of the reference design need usually consider a margin to the conditions that would verify the design premises, but whether there is a need for such margins lies outside the scope of the current document. The term 'withstand' is used in this document in descriptions of load cases on repository components. The statement that a component withstands a particular load means that it upholds its related safety function when exposed to the load in question. For example, if the canister is said to

  12. Contamination, decontamination and radiochemical safety analyses of the RA reactor (Report 1966)

    Maksimovic, Z.

    1966-12-01

    This contract is concerned with development of methods for detection of fission products i the heavy water and quantitative radiochemical analysis for detecting one fission product which enables reliable verification of heavy water contamination by fission products and estimation of contamination level. Qualitative and quantitative radiometry measurements of fission products in water are shown on page 4. Page 6 shows study of contamination and decontamination of water on the laboratory level. Experiments have shown that the majority of fission products was adsorbed on the uranium oxide and that the iodine isotopes are partly in water (non-adsorbed). Gamma spectrometry analyses showed 131 I moves to distillate with the initial quantities of distilled water. decontamination factors compared to the total activity of fission products in distillator and distillate are not higher than ∼10 3 . Decontamination of water contaminated by uranium oxide and fission products in the distillation device of the RA reactor is shown on page 8. Experiments demanded special preparation due to high activity of uranium (1.7 g of uranium irradiated in the reactor for 10 days at neutron flux 1.10 13 n.cm 2 /s. Prior preparations for transport and dissolution of irradiated metal uranium as well as sampling were needed. Distillation was done under lower pressure and temperature to avoid possible contamination of the environment bu fission products and iodine. Decontamination factors are shown in Table. Contamination and decontamination of stainless steel on the laboratory level are described on page 5. It was found that the deposition of activity on the stainless steel plates is inhomogeneous showing that the uranium oxide and fission products are deposited on the rough metal surfaces. According to literature data and our laboratory studies decontamination was done by nitric acid solution (2MHNO 3 ). Since the heavy water system of the RA reactor was made of stainless teel (except the

  13. Biosphere analyses for the safety assessment SR-Site - synthesis and summary of results

    Saetre, Peter [comp.

    2010-12-15

    ice free period between two glaciations. The radionuclide model used in SR-Site has been improved in several important ways since previous safety assessments conducted by SKB. For example, the aquatic and terrestrial ecosystems are handled in the same model, which gives a continuous transition from the sea stage to the lake and terrestrial stages. Transport and accumulation in till (lower regolith) is represented in the model. The uptake by plants is included in the mass-balance, and it is related to biomass growth. Moreover, parameter values including hydrological flows, sedimentation and resuspension rates, biomass growth rates, gas exchange rates, as well as element specific distribution coefficients and concentration rations, were as far as possible based on site data. One endpoint from the simulations with the radionuclide model was the landscape dose conversion factors (LDFs).

  14. Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

    Dickson, T.L.

    1993-07-01

    Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140

  15. Demonstration test on the safety of a cell ventilation system during a hypothetical explosive burning in a fuel reprocessing plant

    Suzuki, Motoe; Nishio, Gunji; Takada, Junichi; Tsukamoto, Michio; Koike, Tadao

    1993-01-01

    To demonstrate the safety of an air ventilation system of cells in a fuel reprocessing plant under a postulated explosive burning caused by solvent fire or by thermal decomposition of nitrated solvent, four types of demonstration tests have been conducted using a large-scale facility simulating a cell ventilation system of an actual reprocessing plant, thus revealing effective mitigation by cell and duct structures on the pressure and temperature pulses generated by explosive burning. In boilover burning tests, solvent fire in a model cell was observed with various sizes of burning surface area as a main parameter, and analysis was performed on the factors dominating the magnitude of boilover burning, revealing that the magnitude strongly depends on accumulated amounts and their ratio of oxygen and solvent vapor present in the cell. In deflagration tests, solid rocket fuel was burned in the cell to simulate the explosive source. The generated pressure and temperature pulses were effectively declined by the cell and duct structures and the integrity of the ventilation system was kept. In blower tests, a centrifugal turbo blower was imposed by a lump of air with a larger flow rate than the rated one by about six times to observe the transient response of the blower fan and motor. It was found that integrity of the blower was kept. In pressure transient tests, compressed air was blown into the cell to induce a mild transient state of fluid dynamics inside the facility, and a variety of data were successfully obtained to be used for the verification and improvement of a computer code. In all the tests, transient overloading of gas caused no damage on HEPA filters, and overloading on the blower motor was avoided either by the slipping of transmission belt or by the acceleration of blower fan rotation during peak flow. (author)

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

    1983-01-01

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

  17. GEOSAF Part II. Demonstration of the operational and long-term safety of geological disposal facilities for radioactive waste. IAEA international intercomparison and harmonization project

    Kumano, Yumiko; Bruno, Gerard [International Atomic Energy Agency, Vienna (Austria). Vienna International Centre; Tichauer, Michael [IRSN, Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France); Hedberg, Bengt [Swedish Radiation Safety Authority, Stockholm (Sweden)

    2015-07-01

    International intercomparison and harmonization projects are one of the mechanisms developed by the IAEA for examining the application and use of safety standards, with a view to ensuring their effectiveness and working towards harmonization of approaches to the safety of radioactive waste management. The IAEA has organized a number of international projects on the safety of radioactive waste management; in particular on the issues related to safety demonstration for radioactive waste management facilities. In 2008, GEOSAF, Demonstration of The Operational and Long-Term Safety of Geological Disposal Facilities for Radioactive Waste, project was initiated. This project was completed in 2011 by delivering a project report focusing on the safety case for geological disposal facilities, a concept that has gained in recent years considerable prominence in the waste management area and is addressed in several international safety standards. During the course of the project, it was recognized that little work was undertaken internationally to develop a common view on the safety approach related to the operational phase of a geological disposal although long-term safety of disposal facility has been discussed for several decades. Upon completion of the first part of the GEOSAF project, it was decided to commence a follow-up project aiming at harmonizing approaches on the safety of geological disposal facilities for radioactive waste through the development of an integrated safety case covering both operational and long-term safety. The new project was named as GEOSAF Part II, which was initiated in 2012 initially as 2-year project, involving regulators and operators. GEOSAF Part II provides a forum to exchange ideas and experience on the development and review of an integrated operational and post-closure safety case for geological disposal facilities. It also aims at providing a platform for knowledge transfer. The project is of particular interest to regulatory

  18. IT-CARES: an interactive tool for case-crossover analyses of electronic medical records for patient safety.

    Caron, Alexandre; Chazard, Emmanuel; Muller, Joris; Perichon, Renaud; Ferret, Laurie; Koutkias, Vassilis; Beuscart, Régis; Beuscart, Jean-Baptiste; Ficheur, Grégoire

    2017-03-01

    The significant risk of adverse events following medical procedures supports a clinical epidemiological approach based on the analyses of collections of electronic medical records. Data analytical tools might help clinical epidemiologists develop more appropriate case-crossover designs for monitoring patient safety. To develop and assess the methodological quality of an interactive tool for use by clinical epidemiologists to systematically design case-crossover analyses of large electronic medical records databases. We developed IT-CARES, an analytical tool implementing case-crossover design, to explore the association between exposures and outcomes. The exposures and outcomes are defined by clinical epidemiologists via lists of codes entered via a user interface screen. We tested IT-CARES on data from the French national inpatient stay database, which documents diagnoses and medical procedures for 170 million inpatient stays between 2007 and 2013. We compared the results of our analysis with reference data from the literature on thromboembolic risk after delivery and bleeding risk after total hip replacement. IT-CARES provides a user interface with 3 columns: (i) the outcome criteria in the left-hand column, (ii) the exposure criteria in the right-hand column, and (iii) the estimated risk (odds ratios, presented in both graphical and tabular formats) in the middle column. The estimated odds ratios were consistent with the reference literature data. IT-CARES may enhance patient safety by facilitating clinical epidemiological studies of adverse events following medical procedures. The tool's usability must be evaluated and improved in further research. © The Author 2016. Published by Oxford University Press on behalf of the American Medical Informatics Association.

  19. Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

    Wulff, W.

    1990-01-01

    A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs

  20. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    Blomquist, C. A. [Argonne National Lab., IL (United States); Ariman, T. [Univ. of Notre Dame, IN (United States); Pierce, R. D.; Pedersen, D. R. [Argonne National Lab., IL (United States)

    1977-07-01

    The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. The cup principal components are: 1. A 38 mm diameter tungsten spike which provides initial fuel quenching and prevents fuel boiling, 2. A 73 mm inside diameter tungsten liner to isolate the support vessel from the molten material high initial temperature, 3. An insulator which is an expedient for extending the experiment time, and 4. An Inconel 625 vessel which provides the structural support to withstand the thermal and pressure stresses. The spike, liner, and insulator are supported by a hemispherical tungsten end cap which fits inside the hemispherical bottom of the support vessel. This vessel is attached to the 316 stainless steel test train with an Inconel 750 wire-formed retaining ring. Thermal analyses were performed with the Argonne-modified version of the general heat transfer code THTB, based on the instantaneous addition of 3200/sup 0/K molten fuel with a decay heat of 9 W/gm and 1920/sup 0/K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The most severe heating occurs when there is no sodium flow outside the cup. For this case the sodium boils (approximately 1200/sup 0/K) and the Inconel vessel and tungsten liner temperatures are approximately 1250/sup 0/K and 2420/sup 0/K, respectively.

  1. Tofacitinib, an oral Janus kinase inhibitor, in patients from Mexico with rheumatoid arthritis: Pooled efficacy and safety analyses from Phase 3 and LTE studies.

    Burgos-Vargas, Ruben; Cardiel, Mario; Xibillé, Daniel; Pacheco-Tena, César; Pascual-Ramos, Virginia; Abud-Mendoza, Carlos; Mahgoub, Ehab; Rahman, Mahboob; Fan, Haiyun; Rojo, Ricardo; García, Erika; Santana, Karina

    2017-05-25

    Tofacitinib is an oral Janus kinase inhibitor for the treatment of rheumatoid arthritis (RA). We characterized efficacy and safety of tofacitinib in Mexican patients from RA Phase 3 and long-term extension (LTE) studies. Data from Mexican patients with RA and an inadequate response to disease-modifying antirheumatic drugs (DMARDs) were taken from four Phase 3 studies (pooled across studies) and one open-label LTE study of tofacitinib. Patients received tofacitinib 5 or 10mg twice daily, adalimumab (one Phase 3 study) or placebo (four Phase 3 studies) as monotherapy or in combination with conventional synthetic DMARDs. Efficacy up to Month 12 (Phase 3) and Month 36 (LTE) was assessed by American College of Rheumatology 20/50/70 response rates, Disease Activity Score (erythrocyte sedimentation rate), and Health Assessment Questionnaire-Disability Index. Safety, including incidence rates (IRs; patients with events/100 patient-years) for adverse events (AEs) of special interest, was assessed throughout the studies. 119 and 212 Mexican patients were included in the Phase 3 and LTE analyses, respectively. Tofacitinib-treated patients in Phase 3 had numerically greater improvements in efficacy responses versus placebo at Month 3. Efficacy was sustained in Phase 3 and LTE studies. IRs for AEs of special interest were similar to those with tofacitinib in the global and Latin American RA populations. In Mexican patients from the tofacitinib global RA program, tofacitinib efficacy was demonstrated up to Month 12 in Phase 3 studies and Month 36 in the LTE study, with a safety profile consistent with tofacitinib global population. Copyright © 2017 Elsevier España, S.L.U. and Sociedad Española de Reumatología y Colegio Mexicano de Reumatología. All rights reserved.

  2. Probabilistic evaluation of scenarios in long-term safety analyses. Results of the project ISIBEL; Probabilistische Bewertung von Szenarien in Langzeitsicherheitsanalysen. Ergebnisse des Vorhabens ISIBEL

    Buhmann, Dieter; Becker, Dirk-Alexander; Laggiard, Eduardo; Ruebel, Andre; Spiessl, Sabine; Wolf, Jens

    2016-07-15

    In the frame of the project ISIBEL deterministic analyses on the radiological consequences of several possible developments of the final repository were performed (VSG: preliminary safety analysis of the site Gorleben). The report describes the probabilistic evaluation of the VSG scenarios using uncertainty and sensitivity analyses. It was shown that probabilistic analyses are important to evaluate the influence of uncertainties. The transfer of the selected scenarios in computational cases and the used modeling parameters are discussed.

  3. TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

    Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

    2010-01-01

    TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

  4. Safety

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  5. Phenomenological analyses and their application to the Defense Waste Processing Facility probabilistic safety analysis accident progression event tree. Revision 1

    Kalinich, D.A.; Thomas, J.K.; Gough, S.T.; Bailey, R.T.; Kearnaghan, D.P.

    1995-01-01

    In the Defense Waste Processing Facility (DWPF) Safety Analysis Reports (SARs) for the Savannah River Site (SRS), risk-based perspectives have been included per US Department of Energy (DOE) Order 5480.23. The NUREG-1150 Level 2/3 Probabilistic Risk Assessment (PRA) methodology was selected as the basis for calculating facility risk. The backbone of this methodology is the generation of an Accident Progression Event Tree (APET), which is solved using the EVNTRE computer code. To support the development of the DWPF APET, deterministic modeling of accident phenomena was necessary. From these analyses, (1) accident progressions were identified for inclusion into the APET; (2) branch point probabilities and any attendant parameters were quantified; and (3) the radionuclide releases to the environment from accidents were determined. The phenomena of interest for accident progressions included explosions, fires, a molten glass spill, and the response of the facility confinement system during such challenges. A variety of methodologies, from hand calculations to large system-model codes, were used in the evaluation of these phenomena

  6. SAFETY

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  7. Safety demonstration tests of air-ventilation system for the postulated explosive burning in a cell of fuel-reprocessing plant

    Takada, Junichi; Suzuki, Motoe; Tukamoto, Michio; Koike, Tadao; Nishio, Gunji

    1995-03-01

    Safety demonstration tests of an explosive burning in a cell in the reprocessing plant has been carried out in JAERI under the auspices of the Science and Technology Agency, to evaluate the safety of an air-ventilation system during the hypothetical explosion. The postulated explosive burning of organic solvent mixed with nitric acid was simulated by solid explosives. The demonstration test was performed using an industrial scale experimental facility simulating to the ventilation system of the large scale reprocessing plant in JAPAN. Propagations of pressure, temperature, and gas velocity through cells and ducts in the ventilation system were measured during the explosive burning under deflagration. Experimental data in this report can be used to evaluate the transport phenomena of radioactive materials in the ventilation system during the explosion, and also to verify computer code CELVA for the safety analysis of ventilation system in the event of explosion accidents. (author)

  8. Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

    Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

    2015-01-01

    The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

  9. 2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

    NONE

    2005-07-01

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

  10. 2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

    NONE

    2005-07-01

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

  11. The long-term safety and performance analyses of the surface disposal facility for the Belgian category a waste at Dessel

    Cool, Wim; Vermarien, Elise; Wacquier, William [ONDRAF/NIRAS Avenue des Arts 14, BE-1210 Bruxelles (Belgium); Perko, Janez [SCK-CEN Boeretang 200, BE-2400 Mol (Belgium)

    2013-07-01

    ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium. This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology. An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal. This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified. (authors)

  12. Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

    2005-03-01

    Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

  13. Pitavastatin demonstrates long-term efficacy, safety and tolerability in elderly patients with primary hypercholesterolaemia or combined (mixed) dyslipidaemia

    Stender, Steen; Budinski, Dragos; Hounslow, Neil

    2013-01-01

    Aims: To assess the long-term efficacy, safety and tolerability of pitavastatin (2 and 4 mg) in elderly patients (≥65 years of age) with primary hypercholesterolaemia or combined (mixed) dyslipidaemia.Design: Patients (n = 545) who had completed a 12-week double-blind comparative study (core study...... Cholesterol Education Program Adult Treatment Plan III (NCEP ATP III) targets for low-density lipoprotein cholesterol (LDL-C) was determined.Results: Of the patients enrolled, 539 received at least one dose of pitavastatin (safety population: men, 45.5%; Caucasian, 99.1%; mean age, 70.3 years; range, 65......-89 years). Only 17% of patients required up-titration to pitavastatin 4 mg. After 60 weeks, NCEP ATP III and EAS targets were attained by 93.8% and 89.0% of patients, respectively. Plasma LDL-C declined by 43.4% and high-density lipoprotein cholesterol increased by 9.6% versus core-study baseline values...

  14. ARAMIS project: a more explicit demonstration of risk control through the use of bow-tie diagrams and the evaluation of safety barrier performance.

    de Dianous, Valérie; Fiévez, Cécile

    2006-03-31

    Over the last two decades a growing interest for risk analysis has been noted in the industries. The ARAMIS project has defined a methodology for risk assessment. This methodology has been built to help the industrialist to demonstrate that they have a sufficient risk control on their site. Risk analysis consists first in the identification of all the major accidents, assuming that safety functions in place are inefficient. This step of identification of the major accidents uses bow-tie diagrams. Secondly, the safety barriers really implemented on the site are taken into account. The barriers are identified on the bow-ties. An evaluation of their performance (response time, efficiency, and level of confidence) is performed to validate that they are relevant for the expected safety function. At last, the evaluation of their probability of failure enables to assess the frequency of occurrence of the accident. The demonstration of the risk control based on a couple gravity/frequency of occurrence is also possible for all the accident scenarios. During the risk analysis, a practical tool called risk graph is used to assess if the number and the reliability of the safety functions for a given cause are sufficient to reach a good risk control.

  15. On groundwater flow modelling in safety analyses of spent fuel disposal. A comparative study with emphasis on boundary conditions

    Jussila, P

    1999-11-01

    Modelling groundwater flow is an essential part of the safety assessment of spent fuel disposal because moving groundwater makes a physical connection between a geological repository and the biosphere. Some of the common approaches to model groundwater flow in bedrock are equivalent porous continuum (EC), stochastic continuum and various fracture network concepts. The actual flow system is complex and measuring data are limited. Multiple distinct approaches and models, alternative scenarios as well as calibration and sensitivity analyses are used to give confidence on the results of the calculations. The correctness and orders of magnitude of results of such complex research can be assessed by comparing them to the results of simplified and robust approaches. The first part of this study is a survey of the objects, contents and methods of the groundwater flow modelling performed in the safety assessment of the spent fuel disposal in Finland and Sweden. The most apparent difference of the Swedish studies compared to the Finnish ones is the approach of using more different models, which is enabled by the more resources available in Sweden. The results of more comprehensive approaches provided by international co-operation are very useful to give perspective to the results obtained in Finland. In the second part of this study, the influence of boundary conditions on the flow fields of a simple 2D model is examined. The assumptions and simplifications in this approach include e.g. the following: (1) the EC model is used, in which the 2-dimensional domain is considered a continuum of equivalent properties without fractures present, (2) the calculations are done for stationary fields, without sources or sinks present in the domain and with a constant density of the groundwater, (3) the repository is represented by an isotropic plate, the hydraulic conductivity of which is given fictitious values, (4) the hydraulic conductivity of rock is supposed to have an exponential

  16. Postauthorization safety surveillance of ADVATE [antihaemophilic factor (recombinant), plasma/albumin-free method] demonstrates efficacy, safety and low-risk for immunogenicity in routine clinical practice.

    Oldenburg, J; Goudemand, J; Valentino, L; Richards, M; Luu, H; Kriukov, A; Gajek, H; Spotts, G; Ewenstein, B

    2010-11-01

      Postauthorization safety surveillance of factor VIII (FVIII) concentrates is essential for assessing rare adverse event incidence. We determined safety and efficacy of ADVATE [antihaemophilic factor (recombinant), plasma/albumin-free method, (rAHF-PFM)] during routine clinical practice. Subjects with differing haemophilia A severities and medical histories were monitored during 12 months of prophylactic and/or on-demand therapy. Among 408 evaluable subjects, 386 (95%) received excellent/good efficacy ratings for all on-demand assessments; the corresponding number for subjects with previous FVIII inhibitors was 36/41 (88%). Among 276 evaluable subjects receiving prophylaxis continuously in the study, 255 (92%) had excellent/good ratings for all prophylactic assessments; the corresponding number for subjects with previous FVIII inhibitors was 41/46 (89%). Efficacy of surgical prophylaxis was excellent/good in 16/16 evaluable procedures. Among previously treated patients (PTPs) with >50 exposure days (EDs) and FVIII≤2%, three (0.75%) developed low-titre inhibitors. Two of these subjects had a positive inhibitor history; thus, the incidence of de novo inhibitor formation in PTPs with FVIII≤2% and no inhibitor history was 1/348 (0.29%; 95% CI, 0.01-1.59%). A PTP with moderate haemophilia developed a low-titre inhibitor. High-titre inhibitors were reported in a PTP with mild disease (following surgery), a previously untreated patient (PUP) with moderate disease (following surgery) and a PUP with severe disease. The favourable benefit/risk profile of rAHF-PFM previously documented in prospective clinical trials has been extended to include a broader range of haemophilia patients, many of whom would have been ineligible for registration studies. © 2010 Blackwell Publishing Ltd.

  17. Experience of RIA safety analyses performance for NPP Temelin core arranged with TVSA-T fuel assemblies

    Kryukov, S.A.; Lizorkin, M.P.

    2010-01-01

    The contents of the presentation are as follows: 1. Definition of categories for initiating events; 2. Acceptance criteria for safety assessment; 3. Main aspects of safety assessment methodology; 4. Main stages of calculation analysis; 5. Interface with other parts of the core design; 6. Codes used for calculation; 6.1 Main performances of code package TIGR-1; 6.2 Main performances of code BIPR-7A; 7. TIGR-1 accounting of design margins in calculation of fuel rod powers; 8. Peculiar features of Instrumentation and Control System for Temelin NPP; 9. Calculations; 10. Checklist of margin data important for reload safety assessment. (P.A.)

  18. Geological boundary conditions for a safety demonstration and verification concept for a HLW repository in claystone in Germany. AnSichT

    Stark, Lena; Bebiolka, Anke; Gerardi, Johannes [Federal Institute for Geosciences and Natural Resources (BGR), Hannover (Germany). Dept. of Underground Space for Storage and Economic Use; and others

    2015-07-01

    Within the framework of the R and D project ''AnSichT'', DBE TECHNOLOGY, BGR and GRS are developing a method to demonstrate the safety of a HLW repository in claystone in Germany. The methodological approach basing on a holistic concept, links the legal and geologic boundary conditions, the disposal and closure concept, the demonstration of barrier integrity, and the long-term analysis of the repository evolution as well. The geologic boundary conditions are specified by the description of the geological situation and generic models, the selection of representative parameters and geoscientific long-term predictions. They form a fundament for the system analysis.

  19. SAFETY

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  20. The safety regulation of small-scale coal mines in China: Analysing the interests and influences of stakeholders

    Song, Xiaoqian; Mu, Xiaoyi

    2013-01-01

    Small scale coal mines (SCMs) have played an important role in China’s energy supply. At the same time, they also suffer from many social, economic, environmental, and safety problems. The Chinese government has made considerable efforts to strengthen the safety regulation of the coal mining industry. Yet, few of these efforts have proven to be very effective. This paper analyzes the interests and influences of key stakeholders in the safety regulation of SCMs, which includes the safety regulator, the local government, the mine owner, and mineworkers. We argue that the effective regulation of coal mine safety must both engage and empower mineworkers. - Highlights: ► Small scale coal mines have played an important role in China's energy supply. ► We analyze the interests and influences of key stakeholders in the safety regulation of small coal mines. ► The mineworkers have the strongest interest but least influence. ► An effective regulation must engage the mineworkers, organize, and empower them.

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

    Schweizer A

    2011-02-01

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

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

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

    2011-01-01

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

  3. Sorption data bases for argillaceous rocks and bentonite for the provisional safety analyses for SGT-E2

    Baeyens, B.; Thoenen, T.; Bradbury, M. H.; Marques Fernandes, M.

    2014-11-01

    In Stage 1 of the Sectoral Plan for Deep Geological Repositories, four rock types have been identified as being suitable host rocks for a radioactive waste repository, namely, Opalinus Clay for a high-level (HLW) and a low- and intermediate-level (L/ILW) repository, and 'Brauner Dogger', Effingen Member and Helvetic Marls for a L/ILW repository. Sorption data bases (SDBs) for all of these host rocks are required for the provisional safety analyses, including all of the bounding porewater and mineralogical composition combinations. In addition, SDBs are needed for the rock formations lying below Opalinus Clay (lower confining units) and for the bentonite backfill in the HLW repository. In some previous work Bradbury et al. (2010) have described a methodology for developing sorption data bases for argillaceous rocks and compacted bentonite. The main factors influencing the sorption in such systems are the phyllosilicate mineral content, particular the 2:1 clay mineral content (illite/smectite/illite-smectite mixed layers) and the water chemistry which determines the radionuclide species in the aqueous phase. The source sorption data were taken predominantly from measurements on illite (or montmorillonite in the case of bentonite) and converted to the defined conditions in each system considered using a series of so called conversion factors to take into account differences in mineralogy, in pH and in radionuclide speciation. Finally, a Lab → Field conversion factor was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions. This methodology to develop sorption data bases has been applied to the selected host rocks, lower confining units and compacted bentonite taking into account the mineralogical and porewater composition ranges defined. Confidence in the validity and correctness of this methodology has been built up through additional studies: (i) sorption values obtained in the manner

  4. Sorption data bases for argillaceous rocks and bentonite for the provisional safety analyses for SGT-E2

    Baeyens, B.; Thoenen, T.; Bradbury, M. H.; Marques Fernandes, M.

    2014-11-15

    In Stage 1 of the Sectoral Plan for Deep Geological Repositories, four rock types have been identified as being suitable host rocks for a radioactive waste repository, namely, Opalinus Clay for a high-level (HLW) and a low- and intermediate-level (L/ILW) repository, and 'Brauner Dogger', Effingen Member and Helvetic Marls for a L/ILW repository. Sorption data bases (SDBs) for all of these host rocks are required for the provisional safety analyses, including all of the bounding porewater and mineralogical composition combinations. In addition, SDBs are needed for the rock formations lying below Opalinus Clay (lower confining units) and for the bentonite backfill in the HLW repository. In some previous work Bradbury et al. (2010) have described a methodology for developing sorption data bases for argillaceous rocks and compacted bentonite. The main factors influencing the sorption in such systems are the phyllosilicate mineral content, particular the 2:1 clay mineral content (illite/smectite/illite-smectite mixed layers) and the water chemistry which determines the radionuclide species in the aqueous phase. The source sorption data were taken predominantly from measurements on illite (or montmorillonite in the case of bentonite) and converted to the defined conditions in each system considered using a series of so called conversion factors to take into account differences in mineralogy, in pH and in radionuclide speciation. Finally, a Lab → Field conversion factor was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions. This methodology to develop sorption data bases has been applied to the selected host rocks, lower confining units and compacted bentonite taking into account the mineralogical and porewater composition ranges defined. Confidence in the validity and correctness of this methodology has been built up through additional studies: (i) sorption values obtained in the manner

  5. A human error taxonomy for analysing healthcare incident reports: assessing reporting culture and its effects on safety perfomance

    Itoh, Kenji; Omata, N.; Andersen, Henning Boje

    2009-01-01

    The present paper reports on a human error taxonomy system developed for healthcare risk management and on its application to evaluating safety performance and reporting culture. The taxonomy comprises dimensions for classifying errors, for performance-shaping factors, and for the maturity...

  6. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  7. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  8. Selection of the situations taken into account for the safety demonstration of a repository in deep geological formations - French regulatory guidance and IPSN modelling experience

    Escalier des Orres, P.; Greneche, D.

    1993-01-01

    A regulatory guidance has been recently set up in France for the safety assessment of radwaste deep geological disposal: the present paper deals with the methodology related to the safety demonstration of such a disposal, particularly the situations to be taken into account to address the potential evolution of the repository under natural or human induced events. This approach, based on a selection of events considered as reasonably envisageable, relies on a reference scenario characterized by a great stability of the geological formation and on hypothetical situations corresponding to the occurrence of random events of natural origin or of conventional nature. The implementation of this methodology within the framework of the IPSN (Protection and Nuclear Safety Institute, CEA) participation in the CEC EVEREST project is addressed. This programme consists in the evaluation of the sensitivity of the radiological consequences associated to deep radwaste disposal systems to the different elements of the performance assessment (scenario characteristics, phenomena, physico-chemical parameters) in three types of geological formations (granite, salt and clay).(author). 11 refs., 3 tabs

  9. Patterns of use and impact of standardised MedDRA query analyses on the safety evaluation and review of new drug and biologics license applications.

    Chang, Lin-Chau; Mahmood, Riaz; Qureshi, Samina; Breder, Christopher D

    2017-01-01

    Standardised MedDRA Queries (SMQs) have been developed since the early 2000's and used by academia, industry, public health, and government sectors for detecting safety signals in adverse event safety databases. The purpose of the present study is to characterize how SMQs are used and the impact in safety analyses for New Drug Application (NDA) and Biologics License Application (BLA) submissions to the United States Food and Drug Administration (USFDA). We used the PharmaPendium database to capture SMQ use in Summary Basis of Approvals (SBoAs) of drugs and biologics approved by the USFDA. Characteristics of the drugs and the SMQ use were employed to evaluate the role of SMQ safety analyses in regulatory decisions and the veracity of signals they revealed. A comprehensive search of the SBoAs yielded 184 regulatory submissions approved from 2006 to 2015. Search strategies more frequently utilized restrictive searches with "narrow terms" to enhance specificity over strategies using "broad terms" to increase sensitivity, while some involved modification of search terms. A majority (59%) of 1290 searches used descriptive statistics, however inferential statistics were utilized in 35% of them. Commentary from reviewers and supervisory staff suggested that a small, yet notable percentage (18%) of 1290 searches supported regulatory decisions. The searches with regulatory impact were found in 73 submissions (40% of the submissions investigated). Most searches (75% of 227 searches) with regulatory implications described how the searches were confirmed, indicating prudence in the decision-making process. SMQs have an increasing role in the presentation and review of safety analysis for NDAs/BLAs and their regulatory reviews. This study suggests that SMQs are best used for screening process, with descriptive statistics, description of SMQ modifications, and systematic verification of cases which is crucial for drawing regulatory conclusions.

  10. Patterns of use and impact of standardised MedDRA query analyses on the safety evaluation and review of new drug and biologics license applications.

    Lin-Chau Chang

    Full Text Available Standardised MedDRA Queries (SMQs have been developed since the early 2000's and used by academia, industry, public health, and government sectors for detecting safety signals in adverse event safety databases. The purpose of the present study is to characterize how SMQs are used and the impact in safety analyses for New Drug Application (NDA and Biologics License Application (BLA submissions to the United States Food and Drug Administration (USFDA.We used the PharmaPendium database to capture SMQ use in Summary Basis of Approvals (SBoAs of drugs and biologics approved by the USFDA. Characteristics of the drugs and the SMQ use were employed to evaluate the role of SMQ safety analyses in regulatory decisions and the veracity of signals they revealed.A comprehensive search of the SBoAs yielded 184 regulatory submissions approved from 2006 to 2015. Search strategies more frequently utilized restrictive searches with "narrow terms" to enhance specificity over strategies using "broad terms" to increase sensitivity, while some involved modification of search terms. A majority (59% of 1290 searches used descriptive statistics, however inferential statistics were utilized in 35% of them. Commentary from reviewers and supervisory staff suggested that a small, yet notable percentage (18% of 1290 searches supported regulatory decisions. The searches with regulatory impact were found in 73 submissions (40% of the submissions investigated. Most searches (75% of 227 searches with regulatory implications described how the searches were confirmed, indicating prudence in the decision-making process.SMQs have an increasing role in the presentation and review of safety analysis for NDAs/BLAs and their regulatory reviews. This study suggests that SMQs are best used for screening process, with descriptive statistics, description of SMQ modifications, and systematic verification of cases which is crucial for drawing regulatory conclusions.

  11. Influence of companion diagnostics on efficacy and safety of targeted anti-cancer drugs: systematic review and meta-analyses.

    Ocana, Alberto; Ethier, Josee-Lyne; Díez-González, Laura; Corrales-Sánchez, Verónica; Srikanthan, Amirrtha; Gascón-Escribano, María J; Templeton, Arnoud J; Vera-Badillo, Francisco; Seruga, Bostjan; Niraula, Saroj; Pandiella, Atanasio; Amir, Eitan

    2015-11-24

    Companion diagnostics aim to identify patients that will respond to targeted therapies, therefore increasing the clinical efficacy of such drugs. Less is known about their influence on safety and tolerability of targeted anti-cancer agents. Randomized trials evaluating targeted agents for solid tumors approved by the US Food and Drug Administration since year 2000 were assessed. Odds ratios (OR) and and 95% confidence intervals (CI) were computed for treatment-related death, treatment-discontinuation related to toxicity and occurrence of any grade 3/4 adverse events (AEs). The 12 most commonly reported individual AEs were also explored. ORs were pooled in a meta-analysis. Analysis comprised 41 trials evaluating 28 targeted agents. Seventeen trials (41%) utilized companion diagnostics. Compared to control groups, targeted drugs in experimental arms were associated with increased odds of treatment discontinuation, grade 3/4 AEs, and toxic death irrespective of whether they utilized companion diagnostics or not. Compared to drugs without available companion diagnostics, agents with companion diagnostics had a lower magnitude of increased odds of treatment discontinuation (OR = 1.12 vs. 1.65, p diagnostics were greatest for diarrhea (OR = 1.29 vs. 2.43, p diagnostics are associated with improved safety, and tolerability. Differences were most marked for gastrointestinal, cutaneous and neurological toxicity.

  12. Post-event reviews: Using a quantitative approach for analysing incident response to demonstrate the value of business continuity programmes and increase planning efficiency.

    Vaidyanathan, Karthik

    2017-01-01

    Business continuity management is often thought of as a proactive planning process for minimising impact from large-scale incidents and disasters. While this is true, and it is critical to plan for the worst, consistently validating plan effectiveness against smaller disruptions can enable an organisation to gain key insights about its business continuity readiness, drive programme improvements, reduce costs and provide an opportunity to quantitatively demonstrate the value of the programme to management. This paper describes a post mortem framework which is used as a continuous improvement mechanism for tracking, reviewing and learning from real-world events at Microsoft Customer Service & Support. This approach was developed and adopted because conducting regular business continuity exercises proved difficult and expensive in a complex and distributed operations environment with high availability requirements. Using a quantitative approach to measure response to incidents, and categorising outcomes based on such responses, enables business continuity teams to provide data-driven insights to leadership, change perceptions of incident root cause, and instil a higher level of confidence towards disaster response readiness and incident management. The scope of the framework discussed here is specific to reviewing and driving improvements from operational incidents. However, the concept can be extended to learning and evolving readiness plans for other types of incidents.

  13. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    Blomquist, C.A.; Ariman, T.; Pierce, R.D.; Pedersen, D.R.

    1977-01-01

    A description of a meltdown cup to be used in the SLSF in-reactor experiments is presented. Thermal analyses have shown that the cup is capable of containing and cooling the postulated quantities of molten fuel and steel. The basic loadings for stress analyses were defined and failure modes were determined. It was shown that both the maximum bending stress and maximum tangential stress in the Inconel vessel are below the material yield stress. Additionally, the axial stress in the Inconel vessel was found to be negligible. The shear stress in the wire-formed retaining ring is much below the maximum shear stress. Therefore, the meltdown cup is capable of performing its required function

  14. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  15. State of the art in establishing computed models of adsorption processes to serve as a basis of radionuclide migration assessment for safety analyses

    Koss, V.

    1991-01-01

    An important point in safety analysis of an underground repository is adsorption of radionuclides in the overlying cover. Adsorption may be judged according to experimental results or to model calculations. Because of the reliability aspired in safety analyses, it is necessary to strengthen experimental results by theoretical calculations. At the time, there is no single thermodynamic model of adsorption to be agreed on. Therefore, this work reviews existing equilibrium models of adsorption. Limitations of the K d -concept and of adsorption-isotherms according to Freundlich and Langmuir are mentioned. The surface ionisation and complexation edl model is explained in full as is the criticism of this model. The application is stressed of simple surface complexation models to adsorption experiments in natural systems as is experimental and modelling work according to systems from Gorleben. Hints are given how to deal with modelling of adsorption related to Gorleben systems in the future. (orig.) [de

  16. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

  17. Safety assessment of historical masonry churches based on pre-assigned kinematic limit analysis, FE limit and pushover analyses

    Milani, Gabriele, E-mail: milani@stru.polimi.it; Valente, Marco, E-mail: milani@stru.polimi.it [Department of Architecture, Built Environment and Construction Engineering (ABC), Politecnico di Milano, Piazza Leonardo da Vinci 32, 20133 Milan (Italy)

    2014-10-06

    This study presents some results of a comprehensive numerical analysis on three masonry churches damaged by the recent Emilia-Romagna (Italy) seismic events occurred in May 2012. The numerical study comprises: (a) pushover analyses conducted with a commercial code, standard nonlinear material models and two different horizontal load distributions; (b) FE kinematic limit analyses performed using a non-commercial software based on a preliminary homogenization of the masonry materials and a subsequent limit analysis with triangular elements and interfaces; (c) kinematic limit analyses conducted in agreement with the Italian code and based on the a-priori assumption of preassigned failure mechanisms, where the masonry material is considered unable to withstand tensile stresses. All models are capable of giving information on the active failure mechanism and the base shear at failure, which, if properly made non-dimensional with the weight of the structure, gives also an indication of the horizontal peak ground acceleration causing the collapse of the church. The results obtained from all three models indicate that the collapse is usually due to the activation of partial mechanisms (apse, façade, lateral walls, etc.). Moreover the horizontal peak ground acceleration associated to the collapse is largely lower than that required in that seismic zone by the Italian code for ordinary buildings. These outcomes highlight that structural upgrading interventions would be extremely beneficial for the considerable reduction of the seismic vulnerability of such kind of historical structures.

  18. Safety assessment of historical masonry churches based on pre-assigned kinematic limit analysis, FE limit and pushover analyses

    Milani, Gabriele; Valente, Marco

    2014-01-01

    This study presents some results of a comprehensive numerical analysis on three masonry churches damaged by the recent Emilia-Romagna (Italy) seismic events occurred in May 2012. The numerical study comprises: (a) pushover analyses conducted with a commercial code, standard nonlinear material models and two different horizontal load distributions; (b) FE kinematic limit analyses performed using a non-commercial software based on a preliminary homogenization of the masonry materials and a subsequent limit analysis with triangular elements and interfaces; (c) kinematic limit analyses conducted in agreement with the Italian code and based on the a-priori assumption of preassigned failure mechanisms, where the masonry material is considered unable to withstand tensile stresses. All models are capable of giving information on the active failure mechanism and the base shear at failure, which, if properly made non-dimensional with the weight of the structure, gives also an indication of the horizontal peak ground acceleration causing the collapse of the church. The results obtained from all three models indicate that the collapse is usually due to the activation of partial mechanisms (apse, façade, lateral walls, etc.). Moreover the horizontal peak ground acceleration associated to the collapse is largely lower than that required in that seismic zone by the Italian code for ordinary buildings. These outcomes highlight that structural upgrading interventions would be extremely beneficial for the considerable reduction of the seismic vulnerability of such kind of historical structures

  19. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  20. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    Hartmann, Tobias

    2013-07-03

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  1. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    Hartmann, Tobias

    2013-01-01

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  2. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  3. Hypnotics and driving safety: meta-analyses of randomized controlled trials applying the on-the-road driving test.

    Verster, Joris C; Veldhuijzen, Dieuwke S; Patat, Alain; Olivier, Berend; Volkerts, Edmund R

    2006-01-01

    Many people who use hypnotics are outpatients and are likely to drive a car the day after drug intake. The purpose of these meta-analyses was to determine whether or not this is safe. Placebo-controlled, randomized, double-blind trials were selected if using the on-the-road driving test to determine driving ability the day following one or two nights of treatment administration. Primary outcome measure of the driving test was the Standard Deviation of Lateral Position (SDLP); i.e., the weaving of the car. Fixed effects model meta-analyses were performed. Effect size (ES) was computed using mean standardized (weighted) difference scores between treatment and corresponding placebo SDLP values. Ten studies, published from 1984 to 2002 (207 subjects), were included in the meta-analyses. The morning following bedtime administration, i.e. 10-11 hours after dosing, significant driving impairment was found for the recommended dose of various benzodiazepine hypnotics (ES=0.42; 95% Confidence Interval (CI)=0.14 to 0.71). Twice the recommended dose impaired driving both in the morning (ES=0.68; CI=0.39 to 0.97) and afternoon, i.e. 16-17 hours after dosing (ES=0.57; CI=0.26 to 0.88). Zopiclone 7.5 mg also impaired driving in the morning (ES=0.89; CI=0.54 to 1.23). Zaleplon (10 and 20 mg) and zolpidem (10 mg) did not affect driving performance the morning after dosing. Following middle-of-the-night administration, significantly impaired driving performance was found for zopiclone 7.5 mg (ES=1.51, CI=0.85 to 2.17), zolpidem 10 mg (ES=0.66, CI=0.13 to 1.19) and zolpidem 20 mg (ES=1.16, CI=0.60 to 1.72). Zaleplon (10 and 20 mg) did not affect driving performance. The analyses show that driving a car the morning following nocturnal treatment with benzodiazepines and zopiclone is unsafe, whereas the recommended dose of zolpidem (10 mg) and zaleplon (10 mg) do not affect driving ability.

  4. Boron analyses in the reactor coolant system of French PWR by acid-base titration ([B]) and ICP-MS (10B atomic %): key to NPP safety

    Jouvet, Fabien; Roux, Sylvie; Carabasse, Stephanie; Felgines, Didier

    2012-09-01

    Boron is widely used by Nuclear Power Plants and especially by EDF Pressurized Water Reactors to ensure the control of the neutron rate in the reactor coolant system and, by this way, the fission reaction. The Boron analysis is thus a major factor of safety which enables operators to guarantee the permanent control of the reactor. Two kinds of analyses carried out by EDF on the Boron species, recently upgraded regarding new method validation standards and developed to enhance the measurement quality by reducing uncertainties, will be discussed in this topic: Acid-Base titration of Boron and Boron isotopic composition by Inductively Coupled Plasma Mass Spectrometer - ICP MS. (authors)

  5. Organic Tank Safety Project: development of a method to measure the equilibrium water content of Hanford organic tank wastes and demonstration of method on actual waste

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1996-09-01

    Some of Hanford's underground waste storage tanks contain Organic- bearing high level wastes that are high priority safety issues because of potentially hazardous chemical reactions of organics with inorganic oxidants in these wastes such as nitrates and nitrites. To ensure continued safe storage of these wastes, Westinghouse Hanford Company has placed affected tanks on the Organic Watch List and manages them under special rules. Because water content has been identified as the most efficient agent for preventing a propagating reaction and is an integral part of the criteria developed to ensure continued safe storage of Hanford's organic-bearing radioactive tank wastes, as part of the Organic Tank Safety Program the Pacific Northwest National Laboratory developed and demonstrated a simple and easily implemented procedure to determine the equilibrium water content of these potentially reactive wastes exposed to the range of water vapor pressures that might be experienced during the wastes' future storage. This work focused on the equilibrium water content and did not investigate the various factors such as at sign ventilation, tank surface area, and waste porosity that control the rate that the waste would come into equilibrium, with either the average Hanford water partial pressure 5.5 torr or other possible water partial pressures

  6. Organic Tank Safety Project: development of a method to measure the equilibrium water content of Hanford organic tank wastes and demonstration of method on actual waste

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1996-09-01

    Some of Hanford`s underground waste storage tanks contain Organic- bearing high level wastes that are high priority safety issues because of potentially hazardous chemical reactions of organics with inorganic oxidants in these wastes such as nitrates and nitrites. To ensure continued safe storage of these wastes, Westinghouse Hanford Company has placed affected tanks on the Organic Watch List and manages them under special rules. Because water content has been identified as the most efficient agent for preventing a propagating reaction and is an integral part of the criteria developed to ensure continued safe storage of Hanford`s organic-bearing radioactive tank wastes, as part of the Organic Tank Safety Program the Pacific Northwest National Laboratory developed and demonstrated a simple and easily implemented procedure to determine the equilibrium water content of these potentially reactive wastes exposed to the range of water vapor pressures that might be experienced during the wastes` future storage. This work focused on the equilibrium water content and did not investigate the various factors such as @ ventilation, tank surface area, and waste porosity that control the rate that the waste would come into equilibrium, with either the average Hanford water partial pressure 5.5 torr or other possible water partial pressures.

  7. Modelling software failures of digital I and C in probabilistic safety analyses based on the TELEPERM registered XS operating experience

    Jockenhoevel-Barttfeld, Mariana; Taurines Andre; Baeckstroem, Ola; Holmberg, Jan-Erik; Porthin, Markus; Tyrvaeinen, Tero

    2015-01-01

    Digital instrumentation and control (I and C) systems appear as upgrades in existing nuclear power plants (NPPs) and in new plant designs. In order to assess the impact of digital system failures, quantifiable reliability models are needed along with data for digital systems that are compatible with existing probabilistic safety assessments (PSA). The paper focuses on the modelling of software failures of digital I and C systems in probabilistic assessments. An analysis of software faults, failures and effects is presented to derive relevant failure modes of system and application software for the PSA. The estimations of software failure probabilities are based on an analysis of the operating experience of TELEPERM registered XS (TXS). For the assessment of application software failures the analysis combines the use of the TXS operating experience at an application function level combined with conservative engineering judgments. Failure probabilities to actuate on demand and of spurious actuation of typical reactor protection application are estimated. Moreover, the paper gives guidelines for the modelling of software failures in the PSA. The strategy presented in this paper is generic and can be applied to different software platforms and their applications.

  8. Labor unions and safety climate: perceived union safety values and retail employee safety outcomes.

    Sinclair, Robert R; Martin, James E; Sears, Lindsay E

    2010-09-01

    Although trade unions have long been recognized as a critical advocate for employee safety and health, safety climate research has not paid much attention to the role unions play in workplace safety. We proposed a multiple constituency model of workplace safety which focused on three central safety stakeholders: top management, ones' immediate supervisor, and the labor union. Safety climate research focuses on management and supervisors as key stakeholders, but has not considered whether employee perceptions about the priority their union places on safety contributes contribute to safety outcomes. We addressed this gap in the literature by investigating unionized retail employee (N=535) perceptions about the extent to which their top management, immediate supervisors, and union valued safety. Confirmatory factor analyses demonstrated that perceived union safety values could be distinguished from measures of safety training, workplace hazards, top management safety values, and supervisor values. Structural equation analyses indicated that union safety values influenced safety outcomes through its association with higher safety motivation, showing a similar effect as that of supervisor safety values. These findings highlight the need for further attention to union-focused measures related to workplace safety as well as further study of retail employees in general. We discuss the practical implications of our findings and identify several directions for future safety research. 2009 Elsevier Ltd. All rights reserved.

  9. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  10. Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

    David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

  11. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    Wagner, J.C.; Parks, C.V.

    2000-01-01

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k inf estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k inf estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration (approx. 2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion (le 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE

  12. SAFETY

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  13. Modeling of a confinement bypass accident with CONSEN, a fast-running code for safety analyses in fusion reactors

    Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it [Sapienza University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Roma (Italy); Giannetti, Fabio [Sapienza University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Roma (Italy); Porfiri, Maria Teresa [ENEA FUS C.R. Frascati, Via Enrico Fermi, 45, 00044 Frascati, Roma (Italy)

    2013-12-15

    Highlights: • The CONSEN code for thermal-hydraulic transients in fusion plants is introduced. • A magnet induced confinement bypass accident in ITER has been simulated. • A comparison with previous MELCOR results for the accident is presented. -- Abstract: The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.

  14. Systematic review of economic analyses in patient safety: a protocol designed to measure development in the scope and quality of evidence.

    Carter, Alexander W; Mandavia, Rishi; Mayer, Erik; Marti, Joachim; Mossialos, Elias; Darzi, Ara

    2017-08-18

    Recent avoidable failures in patient care highlight the ongoing need for evidence to support improvements in patient safety. According to the most recent reviews, there is a dearth of economic evidence related to patient safety. These reviews characterise an evidence gap in terms of the scope and quality of evidence available to support resource allocation decisions. This protocol is designed to update and improve on the reviews previously conducted to determine the extent of methodological progress in economic analyses in patient safety. A broad search strategy with two core themes for original research (excluding opinion pieces and systematic reviews) in 'patient safety' and 'economic analyses' has been developed. Medline, Econlit and National Health Service Economic Evaluation Database bibliographic databases will be searched from January 2007 using a combination of medical subject headings terms and research-derived search terms (see table 1). The method is informed by previous reviews on this topic, published in 2012. Screening, risk of bias assessment (using the Cochrane collaboration tool) and economic evaluation quality assessment (using the Drummond checklist) will be conducted by two independent reviewers, with arbitration by a third reviewer as needed. Studies with a low risk of bias will be assessed using the Drummond checklist. High-quality economic evaluations are those that score >20/35. A qualitative synthesis of evidence will be performed using a data collection tool to capture the study design(s) employed, population(s), setting(s), disease area(s), intervention(s) and outcome(s) studied. Methodological quality scores will be compared with previous reviews where possible. Effect size(s) and estimate uncertainty will be captured and used in a quantitative synthesis of high-quality evidence, where possible. Formal ethical approval is not required as primary data will not be collected. The results will be disseminated through a peer

  15. Safety

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  16. The role of safety analyses in site selection. Some personal observations based on the experience from the Swiss site selection process

    Zuidema, Piet [Nagra, Wettingen (Switzerland)

    2015-07-01

    geological barrier (host rock and confining units); long-term stability (erosion, differential movements, etc.); reliability of geological information (explorability; predictability); technical feasibility (sufficient space for allocating the disposal rooms; depth of repository; rock strength, etc.). For some of these issues, rather detailed quantitative analyses are made (e.g. for erosion). Besides long-term safety, also operational safety is considered. This is done to ensure that suitable sites are chosen for the surface infrastructure (waste acceptance facilities, entrance to access to underground). The main emphasis is on external events (e.g. very severe flooding) that need to be avoided. The involvement of society in the site selection process is also very important. This requires that the scientific information needed (and wanted) by society is delivered in a format understandable to them. This helps society develop an understanding of the question ''why here and not there'' in the siting decision; something that is considered essential to get the necessary support for the siting decision.

  17. The role of safety analyses in site selection. Some personal observations based on the experience from the Swiss site selection process

    Zuidema, Piet

    2015-01-01

    geological barrier (host rock and confining units); long-term stability (erosion, differential movements, etc.); reliability of geological information (explorability; predictability); technical feasibility (sufficient space for allocating the disposal rooms; depth of repository; rock strength, etc.). For some of these issues, rather detailed quantitative analyses are made (e.g. for erosion). Besides long-term safety, also operational safety is considered. This is done to ensure that suitable sites are chosen for the surface infrastructure (waste acceptance facilities, entrance to access to underground). The main emphasis is on external events (e.g. very severe flooding) that need to be avoided. The involvement of society in the site selection process is also very important. This requires that the scientific information needed (and wanted) by society is delivered in a format understandable to them. This helps society develop an understanding of the question ''why here and not there'' in the siting decision; something that is considered essential to get the necessary support for the siting decision.

  18. Safety of high speed ground transportation systems: X2000 US demonstration vehicle dynamics trials, preliminary test report. Report for October 1992-January 1993

    Whitten, B.T.; Kesler, J.K.

    1993-01-01

    The report documents the procedures, events, and results of vehicle dynamic tests carried out on the ASEA-Brown Boveri (ABB) X2000 tilt body trainset in the US between October 1992 and January 1993. These tests, sponsored by Amtrak and supported by the FRA, were conducted to assess the suitability of the X2000 trainset for safe operation at elevated cant deficiencies and speeds in Amtrak's Northeast Corridor under existing track conditions in a revenue service demonstration. The report describes the safety criteria against which the performance of the X2000 test train was examined, the instrumentation used, the test locations, and the track conditions. Preliminary results are presented from tests conducted on Amtrak lines between Philadelphia and Harrisburg, PA, and between Washington DC and New York NY, in which cant deficiencies of 12.5 inches and speeds of 154 mph were reached in a safe and controlled manner. The significance of the results is discussed, and preliminary conclusions and recommendations are presented.

  19. The project ANSICHT. Safety and demonstration methodology for a final repository in clay formations in Germany; Projekt ANSICHT. Sicherheits- und Nachweismethodik fuer ein Endlager im Tongestein in Deutschland. Synthesebericht

    Jobmann, Michael; Bebiolka, Anke; Jahn, Steffen; and others

    2017-03-30

    Based on the status of science and technology and under consideration of international repository concepts the fundamental methodology for safety demonstration for a high-level radioactive waste final repository in clay formations Germany was developed. Basic elements of the safety concept are the geological site description and the geo-scientific long-term prognosis on future performance. Another important section is the closure and sealing concept for the mine shafts. In the frame of the project the fundamental elements were developed and documented for model regions in northern and southern Germany. Three independent safety proofs have to be performed: the demonstration of the geological barrier integrity (clay), the demonstration of the geo-technical barrier system integrity - i.e. closure constructions and backfilling of the shafts, and the radiological demonstration that the radionuclide release in the area is lower than the respective limiting value.

  20. Nuclear Criticality Safety Assessment Using the SCALE Computer Code Package. A demonstration based on an independent review of a real application

    Mennerdahl, Dennis

    1998-06-01

    The purpose of this project was to instruct a young scientist from the Lithuanian Energy Institute (LEI) on how to carry out an independent review of a safety report. In particular, emphasis, was to be put on how to use the personal computer version of the calculation system SCALE 4.3 in this process. Nuclear criticality safety together with radiation shielding from gamma and neutron sources were areas of interest. This report concentrates on nuclear criticality safety aspects while a separate report covers radiation shielding. The application was a proposed storage cask for irradiated fuel assemblies from the Ignalina RBMK reactors in Lithuania. The safety report contained various documents involving many design and safety considerations. A few other documents describing the Ignalina reactors and their operation were available. The time for the project was limited to approximately one month, starting 'clean' with a SCALE 4.3 CD-ROM, a thick safety report and a fast personal computer. The results should be of general interest to Swedish authorities, in particular related to shielding where experience in using advanced computer codes like those available in SCALE is limited. It has been known for many years that criticality safety is very complicated, and that independent reviews are absolutely necessary to reduce the risk from quite common errors in the safety assessments. Several important results were obtained during the project. Concerning use of SCALE 4.3, it was confirmed that a young scientist, without extensive previous experience in the code system, can learn to use essentially all options. During the project, it was obvious that familiarity with personal computers, operating systems (including network system) and office software (word processing, spreadsheet and Internet browser software) saved a lot of time. Some of the Monte Carlo calculations took several hours. Experience is valuable in quickly picking out input or source document errors. Understanding

  1. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 1

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  2. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  3. Final efficacy, immunogenicity, and safety analyses of a nine-valent human papillomavirus vaccine in women aged 16-26 years

    Huh, Warner K; Joura, Elmar A; Giuliano, Anna R

    2017-01-01

    BACKGROUND: Primary analyses of a study in young women aged 16-26 years showed efficacy of the nine-valent human papillomavirus (9vHPV; HPV 6, 11, 16, 18, 31, 33, 45, 52, and 58) vaccine against infections and disease related to HPV 31, 33, 45, 52, and 58, and non-inferior HPV 6, 11, 16, and 18...... antibody responses when compared with quadrivalent HPV (qHPV; HPV 6, 11, 16, and 18) vaccine. We aimed to report efficacy of the 9vHPV vaccine for up to 6 years following first administration and antibody responses over 5 years. METHODS: We undertook this randomised, double-blind, efficacy, immunogenicity......, and safety study of the 9vHPV vaccine study at 105 study sites in 18 countries. Women aged 16-26 years old who were healthy, with no history of abnormal cervical cytology, no previous abnormal cervical biopsy results, and no more than four lifetime sexual partners were randomly assigned (1:1) by central...

  4. Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

    Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

  5. Patient safety climate and worker safety behaviours in acute hospitals in Scotland.

    Agnew, Cakil; Flin, Rhona; Mearns, Kathryn

    2013-06-01

    To obtain a measure of hospital safety climate from a sample of National Health Service (NHS) acute hospitals in Scotland and to test whether these scores were associated with worker safety behaviors, and patient and worker injuries. Data were from 1,866 NHS clinical staff in six Scottish acute hospitals. A Scottish Hospital Safety Questionnaire measured hospital safety climate (Hospital Survey on Patient Safety Culture), worker safety behaviors, and worker and patient injuries. The associations between the hospital safety climate scores and the outcome measures (safety behaviors, worker and patient injury rates) were examined. Hospital safety climate scores were significantly correlated with clinical workers' safety behavior and patient and worker injury measures, although the effect sizes were smaller for the latter. Regression analyses revealed that perceptions of staffing levels and managerial commitment were significant predictors for all the safety outcome measures. Both patient-specific and more generic safety climate items were found to have significant impacts on safety outcome measures. This study demonstrated the influences of different aspects of hospital safety climate on both patient and worker safety outcomes. Moreover, it has been shown that in a hospital setting, a safety climate supporting safer patient care would also help to ensure worker safety. The Scottish Hospital Safety Questionnaire has proved to be a usable method of measuring both hospital safety climate as well as patient and worker safety outcomes. Copyright © 2013 National Safety Council and Elsevier Ltd. Published by Elsevier Ltd. All rights reserved.

  6. Using US EPA’s Chemical Safety for Sustainability’s Comptox Chemistry Dashboard and Tools for Bioactivity, Chemical and Toxicokinetic Modeling Analyses (Course at 2017 ISES Annual Meeting)

    Title: Using US EPA’s Chemical Safety for Sustainability’s Comptox Chemistry Dashboard and Tools for Bioactivity, Chemical and Toxicokinetic Modeling Analyses • Class format: half-day (4 hours) • Course leader(s): Barbara A. Wetmore and Antony J. Williams,...

  7. Application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal. Final report; Anwendung und Weiterentwicklung von Modellen fuer Endlagersicherheitsanalysen auf die Freigabe radioaktiver Stoffe zur Deponierung. Abschlussbericht

    Artmann, Andreas; Larue, Juergen; Seher, Holger; Weiss, Dietmar

    2014-08-15

    The project of application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal is aimed to study the long-term safety using repository-specific simulation programs with respect to radiation exposure for different scenarios. It was supposed to investigate whether the 10 micro Sv criterion can be guaranteed under consideration of human intrusion scenarios. The report covers the following issues: selection and identification of models and codes and the definition of boundary conditions; applicability of conventional repository models for long-term safety analyses; modeling results for the pollutant release and transport and calculation of radiation exposure; determination of the radiation exposure.

  8. Updated safety analysis of ITER

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

    2011-01-01

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

  9. Updated safety analysis of ITER

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

    2011-10-15

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

  10. Comparative analysis of operation and safety of subcritical nuclear systems and innovative critical reactors; Analyse comparative du fonctionnement et de la surete de systemes sous-critiques et de reacteurs critiques innovants

    Bokov, P.M

    2005-05-01

    The main goal of this thesis work is to investigate the role of core subcriticality for safety enhancement of advanced nuclear systems, in particular, molten salt reactors, devoted to both energy production and waste incineration/transmutation. The inherent safety is considered as ultimate goal of this safety improvement. An attempt to apply a systematic approach for the analysis of the subcriticality contribution to inherent properties of hybrid system was performed. The results of this research prove that in many cases the subcriticality may improve radically the safety characteristics of nuclear reactors, and in some configurations it helps to reach the 'absolute' intrinsic safety. In any case, a proper choice of subcriticality level makes all analyzed transients considerably slower and monotonic. It was shown that the weakest point of the independent-source systems with respect to the intrinsic safety is thermohydraulic unprotected transients, while in the case of the coupled-source systems the excess reactivity/current insertion events remain a matter of concern. To overcome these inherent drawbacks a new principle of realization of a coupled sub-critical system (DENNY concept) is proposed. In addition, the ways to remedy some particular safety-related problems with the help of the core sub-criticality are demonstrated. A preliminary safety analysis of the fast-spectrum molten salt reactor (REBUS concept) is also carried out in this thesis work. Finally, the potential of the alternative (to spallation) neutron sources for application in hybrid systems is examined. (author)

  11. Safety analysis for research reactors

    2008-01-01

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

  12. Towards the certification of non-deterministic control systems for safety-critical applications: analysing aviation analogies for possible certification strategies

    Burger, CR

    2011-11-01

    Full Text Available Current certification criteria for safety-critical systems exclude non-deterministic control systems. This paper investigates the feasibility of using human-like monitoring strategies to achieve safe non-deterministic control using multiple...

  13. Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

    Tong Jiejuan; Zhao Jun; Liu Tao; Xue Dazhi

    2012-01-01

    Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited

  14. Complete validation of a unique digestion assay to detect Trichinella larvae in horse meat demonstrates the reliability of this assay for meeting food safety and trade requirements.

    Forbes, L B; Hill, D E; Parker, S; Tessaro, S V; Gamble, H R; Gajadhar, A A

    2008-03-01

    A tissue digestion assay using a double separatory funnel procedure for the detection of Trichinella larvae in horse meat was validated for application in food safety programs and trade. The assay consisted of a pepsin-HCl digestion step to release larvae from muscle tissue and two sequential sedimentation steps in separatory funnels to recover and concentrate larvae for detection with a stereomicroscope. With defined critical control points, the assay was conducted within a quality assurance system compliant with International Organization for Standardization-International Electrotechnical Commission (ISO/IEC) 17025 guidelines. Samples used in the validation were obtained from horses experimentally infected with Trichinella spiralis to obtain a range of muscle larvae densities. One-, 5-, and 10-g samples of infected tissue were combined with 99, 95, and 90 g, respectively, of known negative horse tissue to create a 100-g sample for testing. Samples of 5 and 10 g were more likely to be positive than were 1-g samples when larval densities were less than three larvae per gram (lpg). This difference is important because ingested meat with 1 lpg is considered the threshold for clinical disease in humans. Using a 5-g sample size, all samples containing 1.3 to 2 lpg were detected, and 60 to 100% of samples with infected horse meat containing 0.1 to 0.7 lpg were detected. In this study, the double separatory funnel digestion assay was efficient and reliable for its intended use in food safety and trade. This procedure is the only digestion assay for Trichinella in horse meat that has been validated as consistent and effective at critical levels of sensitivity.

  15. Operational safety at the FFTF

    Baird, Q.L.; Hagan, J.W.; Seeman, S.E.; Baker, S.M.

    1981-02-01

    An extensive operational nuclear safety program has been an integral part of the design, startup, and initial operating phases of the Fast Flux Test Facility (FFTF). During the design and construction of the facility, a program of independent safety overviews and analyses assured the provision of responsible safety margins within the plant, protective systems, and engineered safety features for protection of the public, operating staff, and the facility. The program is continuing through surveillance of operations to verify continued adherence to the established operating envelope and for timely identification of any trends potentially adverse to those margins. Experience from operation of FFTF is being utilized in the development of enhanced operational nuclear safety aids for application in follow-on breeder reactor power systems. The commendable plant and personnel safety experiences of FFTF through its startup and ascension to full power demonstrate the overall effectiveness of the FFTF operational nuclear safety program

  16. German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

    Velkov, K.

    2002-01-01

    Appendix 1: The Safety Analysis Report (S.A.R.) is presented from 3 Handbooks - ECC Handbook (LOCA), Plant Dynamics Handbook (Transients incl. ATWS), and Core Design Handbook. The first one Conceived as Living handbook, Basis for design, catalogue of transients, specifications and licensing. Handbook contains LOCA in primary system, it contains also core damage analysis, and description of codes, description of essential plant data and code input data. The second one consists of Basis for design, commissioning, operation, and catalogue of transients, specifications and licensing, as well as specified operation, disturbed operation, incidents, non-LOCA, SS-procedures and Code description. The third book consists of Reactivity balance and reactivity coefficients, efficiency of shutdown systems. Calculation of burn up cycle, power density distribution, and critical boron concentration. Also Codes used, as SAV79A standard analysis methodology including FASER for nuclear data generation, MEDIUM and PANBOX for static and transient core calculations. Appendix 2: The three TUEV (Technical Inspection Agencies) responsible for the three individual plants of type KONVOI: TUEV Bayern for ISAR-2, TUV-Hanover for KKE, TUEV-Stuttgart for GKN-2 and GRS performed the safety assessment. TUV-Bayern for disturbance and failure of secondary heat sink without loss of coolant (failure of main heat sink, erroneous operation of valves in MS and in FW system, failure of MFW supply), long term LONOP, performance of selected SBLOCA analyses. TUV Hanover for disturbances due to failure of MCPs, short term LONOP, damages of SG tubes incl. SGTR, performance of selected LOCA analyses (blowdown phase of LBLOCA). TUV-Stuttgart for breaks and leaks in MS and FW system with and without leaks in SG tubes. GRS for ATWS, sub-cooling transients due to disturbances on secondary side, initial and boundary conditions for transients with opening of pressurizer valves with and without stuck-open, most of the

  17. Buried Waste Integrated Demonstration Plan

    Kostelnik, K.M.

    1991-12-01

    This document presents the plan of activities for the Buried Waste Integrated Demonstration (BWID) program which supports the environmental restoration (ER) objectives of the Department of Energy (DOE) Complex. Discussed in this plan are the objectives, organization, roles and responsibilities, and the process for implementing and managing BWID. BWID is hosted at the Idaho National Engineering Laboratory (INEL), but involves participants from throughout the DOE Complex, private industry, universities, and the international community. These participants will support, demonstrate, and evaluate a suite of advanced technologies representing a comprehensive remediation system for the effective and efficient remediation of buried waste. The processes for identifying technological needs, screening candidate technologies for applicability and maturity, selecting appropriate technologies for demonstration, field demonstrating, evaluation of results and transferring technologies to environmental restoration programs are also presented. This document further describes the elements of project planning and control that apply to BWID. It addresses the management processes, operating procedures, programmatic and technical objectives, and schedules. Key functions in support of each demonstration such as regulatory coordination, safety analyses, risk evaluations, facility requirements, and data management are presented

  18. Demonstration of safety in Alzheimer's patients for intervention with an anti-hypertensive drug Nilvadipine: results from a 6-week open label study.

    Kennelly, S P

    2012-02-01

    BACKGROUND: Nilvadipine may lower rates of conversion from mild-cognitive impairment to Alzheimer\\'s disease (AD), in hypertensive patients. However, it remains to be determined whether treatment with nilvadipine is safe in AD patients, given the higher incidence of orthostatic hypotension (OH) in this population, who may be more likely to suffer from symptoms associated with the further exaggeration of a drop in BP. OBJECTIVE: The aim of this study was to investigate the safety and tolerability of nilvadipine in AD patients. METHODS: AD patients in the intervention group (n = 56) received nilvadipine 8 mg daily over 6-weeks, compared to the control group (n = 30) who received no intervention. Differences in systolic (SBP) and diastolic (DBP) blood pressure, before and after intervention, was assessed using automated sphygmomanometer readings and ambulatory BP monitors (ABP), and change in OH using a finometer. Reporting of adverse events was monitored throughout the study. RESULTS: There was a significant reduction in the SBP of treated patients compared to non-treated patients but no significant change in DBP. Individuals with higher initial blood pressure (BP) had greater reduction in BP but individuals with normal BP did not experience much change in their BP. While OH was present in 84% of the patients, there was no further drop in BP recorded on active stand studies. There were no significant differences in adverse event reporting between groups. CONCLUSION: Nilvadipine was well tolerated by patients with AD. This study supports further investigation of its efficacy as a potential treatment for AD.

  19. Isoosmolar enemas demonstrate preferential gastrointestinal distribution, safety, and acceptability compared with hyperosmolar and hypoosmolar enemas as a potential delivery vehicle for rectal microbicides.

    Leyva, Francisco J; Bakshi, Rahul P; Fuchs, Edward J; Li, Liye; Caffo, Brian S; Goldsmith, Arthur J; Ventuneac, Ana; Carballo-Diéguez, Alex; Du, Yong; Leal, Jeffrey P; Lee, Linda A; Torbenson, Michael S; Hendrix, Craig W

    2013-11-01

    Rectally applied antiretroviral microbicides for preexposure prophylaxis (PrEP) of HIV infection are currently in development. Since enemas (rectal douches) are commonly used by men who have sex with men prior to receptive anal intercourse, a microbicide enema could enhance PrEP adherence by fitting seamlessly within the usual sexual practices. We assessed the distribution, safety, and acceptability of three enema types-hyperosmolar (Fleet), hypoosmolar (distilled water), and isoosmolar (Normosol-R)-in a crossover design. Nine men received each enema type in random order. Enemas were radiolabeled [(99m)Tc-diethylene triamine pentaacetic acid (DTPA)] to assess enema distribution in the colon using single photon emission computed tomography/computed tomography (SPECT/CT) imaging. Plasma (99m)Tc-DTPA indicated mucosal permeability. Sigmoidoscopic colon tissue biopsies were taken to assess injury as well as tissue penetration of the (99m)Tc-DTPA. Acceptability was assessed after each product use and at the end of the study. SPECT/CT imaging showed that the isoosmolar enema had greater proximal colonic distribution (up to the splenic flexure) and greater luminal and colon tissue concentrations of (99m)Tc-DTPA when compared to the other enemas (pgood with no clear preferences among the three enema types. The isoosmolar enema was superior or similar to the other enemas in all categories and is a good candidate for further development as a rectal microbicide vehicle.

  20. Tested Demonstrations.

    Gilbert, George L.

    1983-01-01

    An apparatus is described in which effects of pressure, volume, and temperature changes on a gas can be observed simultaneously. Includes use of the apparatus in demonstrating Boyle's, Gay-Lussac's, and Charles' Laws, attractive forces, Dalton's Law of Partial pressures, and in illustrating measurable vapor pressures of liquids and some solids.…

  1. Tested Demonstrations.

    Gilbert, George L., Ed.

    1987-01-01

    Describes two demonstrations to illustrate characteristics of substances. Outlines a method to detect the changes in pH levels during the electrolysis of water. Uses water pistols, one filled with methane gas and the other filled with water, to illustrate the differences in these two substances. (TW)

  2. Safety, acceptability, and feasibility of a single-visit approach to cervical-cancer prevention in rural Thailand: a demonstration project.

    Gaffikin, L; Blumenthal, P D; Emerson, M; Limpaphayom, K

    2003-03-08

    To increase screening and treatment coverage, innovative approaches to cervical-cancer prevention are being investigated in rural Thailand. We assessed the value of a single-visit approach combining visual inspection of the cervix with acetic acid wash (VIA) and cryotherapy. 12 trained nurses provided services in mobile (village health centre-based) and static (hospital-based) teams in four districts of Roi-et Province, Thailand. Over 7 months, 5999 women were tested by VIA. If they tested positive, after counselling about the benefits, potential risks, and probable side-effects they were offered cryotherapy. Data measuring safety, acceptability, feasibility, and effort to implement the programme were gathered. The VIA test-positive rate was 13.3% (798/5999), and 98.5% (609/618) of those eligible accepted immediate treatment. Overall, 756 women received cryotherapy, 629 (83.2%) of whom returned for their first follow-up visit. No major complications were recorded, and 33 (4.4%) of those treated returned for a perceived problem. Only 17 (2.2%) of the treated women needed clinical management other than reassurance about side-effects. Both VIA and cryotherapy were highly acceptable to the patients (over 95% expressed satisfaction with their experience). At their 1-year visit, the squamocolumnar junction was visible to the nurses, and the VIA test-negative rate was 94.3%. A single-visit approach with VIA and cryotherapy seems to be safe, acceptable, and feasible in rural Thailand, and is a potentially efficient method of cervical-cancer prevention in such settings.

  3. A New Safety Concern for Glaucoma Treatment Demonstrated by Mass Spectrometry Imaging of Benzalkonium Chloride Distribution in the Eye, an Experimental Study in Rabbits

    Brignole-Baudouin, Françoise; Desbenoit, Nicolas; Hamm, Gregory; Liang, Hong; Both, Jean-Pierre; Brunelle, Alain; Fournier, Isabelle; Guerineau, Vincent; Legouffe, Raphael; Stauber, Jonathan; Touboul, David; Wisztorski, Maxence; Salzet, Michel; Laprevote, Olivier; Baudouin, Christophe

    2012-01-01

    We investigated in a rabbit model, the eye distribution of topically instilled benzalkonium_(BAK) chloride a commonly used preservative in eye drops using mass spectrometry imaging. Three groups of three New Zealand rabbits each were used: a control one without instillation, one receiving 0.01%BAK twice a day for 5 months and one with 0.2%BAK one drop a day for 1 month. After sacrifice, eyes were embedded and frozen in tragacanth gum. Serial cryosections were alternately deposited on glass slides for histological (hematoxylin-eosin staining) and immunohistological controls (CD45, RLA-DR and vimentin for inflammatory cell infiltration as well as vimentin for Müller glial cell activation) and ITO or stainless steel plates for MSI experiments using Matrix-assisted laser desorption ionization time-of-flight. The MSI results were confirmed by a round-robin study on several adjacent sections conducted in two different laboratories using different sample preparation methods, mass spectrometers and data analysis softwares. BAK was shown to penetrate healthy eyes even after a short duration and was not only detected on the ocular surface structures, but also in deeper tissues, especially in sensitive areas involved in glaucoma pathophysiology, such as the trabecular meshwork and the optic nerve areas, as confirmed by images with histological stainings. CD45-, RLA-DR- and vimentin-positive cells increased in treated eyes. Vimentin was found only in the inner layer of retina in normal eyes and increased in all retinal layers in treated eyes, confirming an activation response to a cell stress. This ocular toxicological study confirms the presence of BAK preservative in ocular surface structures as well as in deeper structures involved in glaucoma disease. The inflammatory cell infiltration and Müller glial cell activation confirmed the deleterious effect of BAK. Although these results were obtained in animals, they highlight the importance of the safety-first principle for

  4. A new safety concern for glaucoma treatment demonstrated by mass spectrometry imaging of benzalkonium chloride distribution in the eye, an experimental study in rabbits.

    Brignole-Baudouin, Françoise; Desbenoit, Nicolas; Hamm, Gregory; Liang, Hong; Both, Jean-Pierre; Brunelle, Alain; Fournier, Isabelle; Guerineau, Vincent; Legouffe, Raphael; Stauber, Jonathan; Touboul, David; Wisztorski, Maxence; Salzet, Michel; Laprevote, Olivier; Baudouin, Christophe

    2012-01-01

    We investigated in a rabbit model, the eye distribution of topically instilled benzalkonium_(BAK) chloride a commonly used preservative in eye drops using mass spectrometry imaging. Three groups of three New Zealand rabbits each were used: a control one without instillation, one receiving 0.01%BAK twice a day for 5 months and one with 0.2%BAK one drop a day for 1 month. After sacrifice, eyes were embedded and frozen in tragacanth gum. Serial cryosections were alternately deposited on glass slides for histological (hematoxylin-eosin staining) and immunohistological controls (CD45, RLA-DR and vimentin for inflammatory cell infiltration as well as vimentin for Müller glial cell activation) and ITO or stainless steel plates for MSI experiments using Matrix-assisted laser desorption ionization time-of-flight. The MSI results were confirmed by a round-robin study on several adjacent sections conducted in two different laboratories using different sample preparation methods, mass spectrometers and data analysis softwares. BAK was shown to penetrate healthy eyes even after a short duration and was not only detected on the ocular surface structures, but also in deeper tissues, especially in sensitive areas involved in glaucoma pathophysiology, such as the trabecular meshwork and the optic nerve areas, as confirmed by images with histological stainings. CD45-, RLA-DR- and vimentin-positive cells increased in treated eyes. Vimentin was found only in the inner layer of retina in normal eyes and increased in all retinal layers in treated eyes, confirming an activation response to a cell stress. This ocular toxicological study confirms the presence of BAK preservative in ocular surface structures as well as in deeper structures involved in glaucoma disease. The inflammatory cell infiltration and Müller glial cell activation confirmed the deleterious effect of BAK. Although these results were obtained in animals, they highlight the importance of the safety-first principle for

  5. A new safety concern for glaucoma treatment demonstrated by mass spectrometry imaging of benzalkonium chloride distribution in the eye, an experimental study in rabbits.

    Françoise Brignole-Baudouin

    Full Text Available We investigated in a rabbit model, the eye distribution of topically instilled benzalkonium_(BAK chloride a commonly used preservative in eye drops using mass spectrometry imaging. Three groups of three New Zealand rabbits each were used: a control one without instillation, one receiving 0.01%BAK twice a day for 5 months and one with 0.2%BAK one drop a day for 1 month. After sacrifice, eyes were embedded and frozen in tragacanth gum. Serial cryosections were alternately deposited on glass slides for histological (hematoxylin-eosin staining and immunohistological controls (CD45, RLA-DR and vimentin for inflammatory cell infiltration as well as vimentin for Müller glial cell activation and ITO or stainless steel plates for MSI experiments using Matrix-assisted laser desorption ionization time-of-flight. The MSI results were confirmed by a round-robin study on several adjacent sections conducted in two different laboratories using different sample preparation methods, mass spectrometers and data analysis softwares. BAK was shown to penetrate healthy eyes even after a short duration and was not only detected on the ocular surface structures, but also in deeper tissues, especially in sensitive areas involved in glaucoma pathophysiology, such as the trabecular meshwork and the optic nerve areas, as confirmed by images with histological stainings. CD45-, RLA-DR- and vimentin-positive cells increased in treated eyes. Vimentin was found only in the inner layer of retina in normal eyes and increased in all retinal layers in treated eyes, confirming an activation response to a cell stress. This ocular toxicological study confirms the presence of BAK preservative in ocular surface structures as well as in deeper structures involved in glaucoma disease. The inflammatory cell infiltration and Müller glial cell activation confirmed the deleterious effect of BAK. Although these results were obtained in animals, they highlight the importance of the safety

  6. Studies to demonstrate the adequacy of testing results of the qualification tests for the actuator of main steam safety relive valves (MSSRV) in an advanced boiling water reactor (ABWR)

    Gou, P.F.; Patel, R.; Curran, G.; Henrie, D.; Solorzano, E.

    2005-01-01

    This paper presents several studies performed to demonstrate that the testing results from the qualification tests for the actuator of the Main Steam Safety Relief Valves (MSSRV; also called SRV in this paper) in GE's Advanced Boiling Water Reactor (ABWR) are in compliance with the qualification guidelines stipulated in the applicable IEEE standards. The safety-related function of the MSSRV is to relieve pressure in order to protect the reactor pressure vessel from over-pressurization condition during normal operation and design basis events. In order to perform this function, the SRV must actuate at a given set pressure while maintaining the pressure and structural integrity of the SRV. The valves are provided with an electro-pneumatic actuator assembly that opens the valve upon receipt of an automatic or manually initiated electric signal to allow depressurization of the reactor pressure vessel (RPV). To assure the SRV can perform its intended safety related functions properly, qualification tests are needed in addition to analysis, to demonstrate that the SRV can withstand the specified environmental, dynamic and seismic design basis conditions without impairing its safety related function throughout their installed life under the design conditions including postulated design basis events such as OBE loads and Faulted (SSE) events. The guidelines used for the test methods, procedures and acceptance criteria for the qualification tests are established in IEEE std 344-1987 and IEEE std 382-1985. In the qualification tests, the specimen consists of the actuator, control valve assembly, limit switches, and limit switch support structure. During the functional, dynamic and seismic tests, the test specimen was mounted on a SRV. Qualification of safety related equipment to meet the guidelines of the IEEE standards is typically a two-step process: 1) environmental aging and 2) design basis events qualification. The purpose of the first step is to put the equipment in an

  7. Fuel pin transient behavior technology applied to safety analyses. Presentation to AEC Regulatory Staff 4th Regulatory Briefing on safety technology, Washington, D.C., November 19--20, 1974

    1974-11-01

    Information is presented concerning LMFBR fuel pin performance requirements and evaluation; fuels behavior codes with safety interfaces; performance evaluations; ex-reactor materials and simulation tests; models for fuel pin failure; and summary of continuing fuels technology tasks. (DCC)

  8. The business case for safety and health at work : cost-benefit analyses of interventions in small and medium-sized enterprises

    Targoutzidis, A.; Koukoulaki, T.; Schmitz-Felten, E.; Kuhl, K.; Oude Hengel, K.M.; Rijken, E.; Broek K. van den; Kluser, R.

    2014-01-01

    This report examines the economic aspects of occupational safety and health (OSH) interventions in small and medium-sized businesses (SMEs). First, case studies in the existing literature were identified and examined. Second, 13 new case studies on OSH initiatives in European SMEs were developed,

  9. Review of the Methods to Obtain Paediatric Drug Safety Information: Spontaneous Reporting and Healthcare Databases, Active Surveillance Programmes, Systematic Reviews and Meta-analyses

    Gentili, Marta; Pozzi, Marco; Peeters, Gabrielle; Radice, Sonia; Carnovale, Carla

    2018-02-06

    Knowledge of drugs safety collected during the pre-marketing phase is inevitably limited because the randomized clinical trials (RCTs) are rarely designed to evaluate safety. The small and selective groups of enrolled individuals and the limited duration of trials may hamper the ability to characterize fully the safety profiles of drugs. Additionally, information about rare adverse drug reactions (ADRs) in special groups is often incomplete or not available for most of the drugs commonly used in the daily clinical practice. In the paediatric setting several highimpact safety issues have emerged. Hence, in recent years, there has been a call for improved post-marketing pharmacoepidemiological studies, in which cohorts of patients are monitored for sufficient time in order to determine the precise risk-benefit ratio. In this review, we discuss the current available strategies enhancing the post-marketing monitoring activities of the drugs in the paediatric setting and define criteria whereby they can provide valuable information to improve the management of therapy in daily clinical practice including both safety and efficacy aspects. The strategies we cover include the signal detection using international pharmacovigilance and/or healthcare databases, the promotion of active surveillance initiatives which can generate complete, informative data sets for the signal detection and systematic review/meta-analysis. Together, these methods provide a comprehensive picture of causality and risk improving the management of therapy in a paediatric setting and they should be considered as a unique tool to be integrated with post-marketing activities. Copyright© Bentham Science Publishers; For any queries, please email at epub@benthamscience.org.

  10. Safety and tolerability of extended-release niacin-laropiprant: Pooled analyses for 11,310 patients in 12 controlled clinical trials.

    McKenney, James; Bays, Harold; Gleim, Gilbert; Mitchel, Yale; Kuznetsova, Olga; Sapre, Aditi; Sirah, Waheeda; Maccubbin, Darbie

    2015-01-01

    The Heart Protection Study 2-Treatment of HDL to Reduce the Incidence of Vascular Events (HPS2-THRIVE) showed that adding extended-release niacin-laropiprant (ERN-LRPT) to statin provided no incremental cardiovascular benefit vs placebo (PBO). ERN-LRPT was also associated with an excess of serious adverse experiences (AEs), some of which were unexpected (infections and bleeding). These findings led to the withdrawal of ERN-LRPT from all markets. We examined the safety profile of ERN-LRPT vs the comparators ERN alone and statins in the ERN-LRPT development program to assess whether similar safety signals were observed to those seen in HPS-THRIVE and whether these might be attributed to ERN or LRPT. Postrandomization safety data from 12 clinical studies, 12 to 52 weeks in duration and involving 11,310 patients, were analyzed across 3 treatments: (1) ERN-LRPT; (2) ERN-NSP (ERN, Merck & Co, Inc or Niaspan [NSP], Abbott Laboratories); and (3) statin-PBO (statin or PBO). The safety profiles of ERN-LRPT and ERN-NSP were similar, except for less flushing with ERN-LRPT. Nonflushing AEs reported more frequently with ERN-LRPT or ERN-NSP than with statin-PBO were mostly nonserious and typical of niacin (nausea, diarrhea, and increased blood glucose). There was no evidence for an increased risk of serious AEs related to diabetes, muscle, infection, or bleeding. Pooled data from 11,310 patients revealed that, except for reduced flushing, the safety profile of ERN-LRPT was similar to that of ERN-NSP; LRPT did not appear to adversely affect the side-effect profile of ERN. The inability to replicate the unexpected AE findings in HPS2-THRIVE could be because of the smaller sample size and substantially shorter duration of these studies. Copyright © 2015 National Lipid Association. Published by Elsevier Inc. All rights reserved.

  11. Effective diffusion coefficients and porosity values for argillaceous rocks and bentonite: measured and estimated values for the provisional safety analyses for SGT-E2

    Van Loon, L.R.

    2014-11-01

    In Stage 2 of the Sectoral Plan for Deep Geological Repositories, safety analyses have to be performed. Geochemical parameters describing the transport and retardation of radionuclides in the argillaceous rocks considered and in compacted bentonite are required. In the present report, diffusion parameters for all clay host rocks, confining units and compacted bentonite are derived. Diffusion of tritiated water (HTO), "3"6Cl"- and "2"2Na"+ was studied. The measurements gave values for effective diffusion coefficients (D_e) and diffusion accessible porosities. The general observed trend "N"aD_e > "H"T"OD_e > "C"lD_e is in agreement with the expected behaviour of the three species in clay materials: ion exchanging cations show an enhanced mobility due to surface diffusion effects and anions are slowed down due to anion exclusion. Due to the negatively charged clay surfaces, anionic species are repelled from these surfaces resulting in an accessible porosity that is smaller than the total porosity as measured with HTO. The effect of porewater composition on the diffusion of HTO, "3"6Cl"- and "2"2Na"+ in Opalinus Clay was investigated. For ionic strength (IS) values between 0.17 M and 1.07 M, no significant effect on D_e could be observed. In the case of "3"6Cl"-, no effect on the accessible porosity was observed. The anion diffusion accessible porosity equals 50-60 % of the total porosity, independent on the ionic strength of the porewater. The diffusion parameters were measured on sedimentary rocks such as chalk, clay and limestone rocks. All data could be described by one single modified version of Archie's relation (extended Archie's relation). For values of porosity greater than about 0.1, the classical Archie's relation was valid. For values smaller than 0.1, the data deviated from the classical Archie's relation; this can be explained by additional changes of tortuosity with porosity values. At high porosity values (low density rocks), the microfabric of the clay

  12. Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

    Serdula, K.J.

    1988-01-01

    Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

  13. ITER safety

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  14. 48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Focus session: Uncertainty analyses in reactor core simulations

    Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

    2017-12-15

    The supplementation of reactor simulations by uncertainty analyses is becoming increasingly important internationally due to the fact that the reliability of simulation calculations can be significantly increased by the quantification of uncertainties in comparison to the use of so-called conservative methods (BEPU- ''Best-Estimate plus Uncertainties''). While systematic uncertainty analyses for thermo-hydraulic calculations have been performed routinely for a long time, methods for taking into account uncertainties in nuclear data, which are the basis for neutron transport calculations, are under development. The Focus Session Uncertainty Analyses in Reactor Core Simulations was intended to provide an overview of international research and development with respect to supplementing reactor core simulations with uncertainty and sensitivity analyses, in research institutes as well as within the nuclear industry. The presented analyses not only focused on light water reactors, but also on advanced reactor systems. Particular emphasis was put on international benchmarks in the field. The session was chaired by Winfried Zwermann (Gesellschaft fuer Anlagen- und Reaktorsicherheit).

  15. Adaptation of computer code ALMOD 3.4 for safety analyses of Westighouse type NPPs and calculation of main feedwater loss

    Kordis, I.; Jerele, A.; Brajak, F.

    1986-01-01

    The paper presents theoretical foundations of ALMOD 3.4 code and modification done in order to adjust the model to westinghouse type NPP. test cases for verification of added modules functioning were done and loss of main feedwater (FW) transient at nominal power was analysed. (author)

  16. Safety; Avertissement

    NONE

    2001-07-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  17. Sorption data base for the cementitious near-field of L/ILW and ILW repositories for provisional safety analyses for SGT-E2

    Wieland, E.

    2014-11-01

    The near-field of the planned Swiss repositories for low- and intermediate-level waste (L/ILW) and long-lived intermediate-level waste (ILW) consists of large quantities of cementitious materials. Hardened cement paste (HCP) is considered to be the most important sorbing material present in the near-field of L/ILW and ILW repositories. Interaction of radionuclides with HCP represents the most important mechanism retarding their migration from the near-field into the host rock. This report describes a cement sorption data base (SDB) for the safety-relevant radionuclides in the waste that will be disposed of in the L/ILW and ILW repositories. The current update on sorption values for radionuclides should be read in conjunction with the earlier SDBs CEM-94, CEM-97 and CEM-02. Sorption values have been selected based on procedures reported in these earlier SDBs. The values are revised if corresponding new information and/or data are available. The basic information results from a survey of sorption studies published between 2002 and 2013. The sorption values recommended in this report have either been selected from in-house experimental studies or from literature data, and they were further assessed with a view to the sorption values recently published in the framework of the safety analysis for the planned near surface disposal facility in Belgium. The report summarizes the sorption properties of HCP and compiles sorption values for safety-relevant radionuclides and low-molecular weight organic molecules on undisturbed and degraded HCP. A list of the safety-relevant radionuclides is provided. The radionuclide inventories are determined by the waste streams to be disposed of in the L/ILW and ILW repositories. Information on the elemental and mineral composition of HCP was obtained from hydration studies. The concentrations of the most important impurity elements in cement were obtained from dissolution studies on HCP. Particular emphasis is placed on summarizing our

  18. LFR safety approach and main ELFR safety analysis results

    Bubelis, E.; Schikorr, M.; Frogheri, M.; Mansani, L.; Bandini, G.; Burgazzi, L.; Mikityuk, K.; Zhang, Y.; Lo Frano, R.; Forgione, N.

    2013-01-01

    LFR safety approach: → A global safety approach for the LFR reference plant has been assessed and the safety analyses methodology has been developed. → LFR follows the general guidelines of the Generation IV safety concept recommendations. Thus, improved safety and higher reliability are recognized as an essential priority. → The fundamental safety objectives and the Defence-in-Depth (DiD) approach, as described by IAEA Safety Guides, have been preserved. → The recommendations of the Risk and Safety Working Group (RSWG) of GEN-IV IF has been taken into account: • safety is to be “built-in” in the fundamental design rather than “added on”; • full implementation of the Defence-in-Depth principles in a manner that is demonstrably exhaustive, progressive, tolerant, forgiving and well-balanced; • “risk-informed” approach - deterministic approach complemented with a probabilistic one; • adoption of an integrated methodology that can be used to evaluate and document the safety of Gen IV nuclear systems - ISAM. In particular the OPT tool is the fundamental methodology used throughout the design process

  19. Fabrication of 3-methoxyphenol sensor based on Fe3O4 decorated carbon nanotube nanocomposites for environmental safety: Real sample analyses.

    Mohammed M Rahman

    Full Text Available Iron oxide ornamented carbon nanotube nanocomposites (Fe3O4.CNT NCs were prepared by a wet-chemical process in basic means. The optical, morphological, and structural characterizations of Fe3O4.CNT NCs were performed using FTIR, UV/Vis., FESEM, TEM; XEDS, XPS, and XRD respectively. Flat GCE had been fabricated with a thin-layer of NCs using a coating binding agent. It was performed for the chemical sensor development by a dependable I-V technique. Among all interfering analytes, 3-methoxyphenol (3-MP was selective towards the fabricated sensor. Increased electrochemical performances for example elevated sensitivity, linear dynamic range (LDR and continuing steadiness towards selective 3-MP had been observed with chemical sensor. The calibration graph found linear (R2 = 0.9340 in a wide range of 3-MP concentration (90.0 pM ~ 90.0 mM. The limit of detection and sensitivity were considered as 1.0 pM and 9×10-4 μAμM-1cm-2 respectively. The prepared of Fe3O4.CNT NCs by a wet-chemical progression is an interesting route for the development of hazardous phenolic sensor based on nanocomposite materials. It is also recommended that 3-MP sensor is exhibited a promising performances based on Fe3O4.CNT NCs by a facile I-V method for the significant applications of toxic chemicals for the safety of environmental and health-care fields.

  20. Development of the status of W and T for the realization of a long-term safety demonstration for the final repository using the examples VSG and Konrad. Report on the Working package 2. Review and development of safety-related assessments of disposal facilities of wastes with negligible heat generation; development and provision of the necessary set of tools using the example of the final repository Konrad

    Larue, Juergen; Fischer-Appelt, Klaus; Hartwig-Thurat, Eva

    2015-09-01

    In the research project on the ''Review and development of safety-related assessments of disposal facilities with negligible heat generation; development and provision of the necessary set of tools, using the example of the Konrad disposal facility'' (3612R03410), the state of the art in science and technology of the safety-related assessments and sets of tools for building a safety case was examined. The reports pertaining to the two work packages described the further development of the methodology for accident analyses (WP 1) and of building a safety case (WP 2); also, comparisons were drawn on a national and international scale with the methods applied in the licensing procedure of the Konrad disposal facility. A safety case as well as its underlying analyses and methods always has to be brought up to date with the development of the state of the art in science and technology. In Germany, two safety cases regarding the long-term safety of disposal facilities have been prepared. These are the licensing documentation for the Konrad disposal facility in the year 1990 and the research project regarding the preliminary safety case for the Gorleben site (Vorlaeufige Sicherheitsanalyse Gorleben - VSG) in the year 2013, both reflecting the state of development of building a safety case at the respective time. Comparing the two above-mentioned examples of safety cases and taking recent international recommendations and national regulations into account, this report on Work Package 2 presents the development of the international state of the art in science and technology. This has been done by summarising the essential differences and similarities of each element of the safety case for the Konrad disposal facility on the one hand and the VSG and the international status on the other hand.

  1. Safety and efficacy of first-line bevacizumab combination therapy in Chinese population with advanced non-squamous NSCLC: data of subgroup analyses from MO19390 (SAiL) study.

    Zhou, C C; Bai, C X; Guan, Z Z; Jiang, G L; Shi, Y K; Wang, M Z; Wu, Y L; Zhang, Y P; Zhu, Y Z

    2014-05-01

    Bevacizumab is a monoclonal antibody with high antitumor activity against malignant diseases. Previous studies have demonstrated the efficacy of first-line bevacizumab combination therapy in advanced, non-squamous non-small cell lung cancer (NS-NSCLC). SAiL (MO19390), an open-label, multicenter, single-arm study, evaluated the safety and efficacy of first-line bevacizumab-based treatment in clinical practice. This report presents the results of a subgroup analysis of Chinese patients enrolled in SAiL. Chemo-naive Chinese patients with locally advanced, metastatic or recurrent NSCLC were randomized to receive Bev 15 mg/kg every 3 weeks plus carboplatin + paclitaxel for maximum of six cycles, followed by single-agent bevacizumab until disease progression. The primary endpoint was safety. Secondary endpoints included time to progression and overall survival. The Chinese intent-to-treat (ITT) population consists of 198 Chinese patients, among whom 107 (54 %) were non-smokers and 90 (45.5 %) were female. The median cycle of bevacizumab administration was 10 and median duration of bevacizumab treatment was 29.5 weeks. Only eight cases of severe adverse events were observed in the study, which were deemed to be related to bevacizumab. The incidence of AEs over grade 3 in Chinese ITT patients was generally low (SAiL study. No new safety signals were reported.

  2. FY 1998 annual report on the demonstration tests for establishing load concentration controlling systems. Survey on safety of commercial systems; Fuka shuchu seigyo system kakuritsu jissho shiken 1998 nendo kenkyu hokokusho. Jjitsuyo system anzensei chosa

    NONE

    1999-03-01

    The demonstration tests are being conducted for establishing load concentration controlling systems, which directly or indirectly control load devices in residential power consumers or the like from a power supplier, as one of the DSM measures. This project is aimed at survey on the systems which support general residential consumers or the like to adequately control loads indirectly, and at clarification of technical essentials the system should have when it is actually put in service and the safety rules to be observed, thereby contributing eventual commercialization of the load concentration controlling systems. The field test results indicate that functions of a monitor set in a domestic consumer can be well operated even by inexperienced persons in handling machines, when they have some experiences. Reliability of a monitoring/controlling device, set in a domestic consumer on a trial basis, can be secured effectively by addition of an LC circuit and changing the modulation mode to FSK. The devices developed on a trial basis are found to be well serviceable for the demonstration tests. The best method for communication with the monitoring/controlling device for electric appliances in a domestic consumer is communication via a power transmission line. (NEDO)

  3. Divergent effects of transformational and passive leadership on employee safety.

    Kelloway, E Kevin; Mullen, Jane; Francis, Lori

    2006-01-01

    The authors concurrently examined the impact of safety-specific transformational leadership and safety-specific passive leadership on safety outcomes. First, the authors demonstrated via confirmatory factor analysis that safety-specific transformational leadership and safety-specific passive leadership are empirically distinct constructs. Second, using hierarchical regression, the authors illustrated, contrary to a stated corollary of transformational leadership theory (B. M. Bass, 1997), that passive leadership contributes incrementally to the prediction of organizationally relevant outcomes, in this case safety-related variables, beyond transformational leadership alone. Third, further analyses via structural equation modeling showed that both transformational and passive leadership have opposite effects on safety climate and safety consciousness, and these variables, in turn, predict safety events and injuries. Implications for research and application are discussed. Copyright 2006 APA.

  4. Efficacy and safety of palbociclib in combination with letrozole as first-line treatment of ER-positive, HER2-negative, advanced breast cancer: expanded analyses of subgroups from the randomized pivotal trial PALOMA-1/TRIO-18.

    Finn, Richard S; Crown, John P; Ettl, Johannes; Schmidt, Marcus; Bondarenko, Igor M; Lang, Istvan; Pinter, Tamas; Boer, Katalin; Patel, Ravindranath; Randolph, Sophia; Kim, Sindy T; Huang, Xin; Schnell, Patrick; Nadanaciva, Sashi; Bartlett, Cynthia Huang; Slamon, Dennis J

    2016-06-28

    Palbociclib is an oral small-molecule inhibitor of cyclin-dependent kinases 4 and 6. In the randomized, open-label, phase II PALOMA-1/TRIO-18 trial, palbociclib in combination with letrozole improved progression-free survival (PFS) compared with letrozole alone as first-line treatment of estrogen receptor (ER)-positive, human epidermal growth factor receptor 2 (HER2)-negative, advanced breast cancer (20.2 months versus 10.2 months; hazard ratio (HR) = 0.488, 95 % confidence interval (CI) 0.319-0.748; one-sided p = 0.0004). Grade 3-4 neutropenia was the most common adverse event (AE) in the palbociclib + letrozole arm. We now present efficacy and safety analyses based on several specific patient and tumor characteristics, and present in detail the clinical patterns of neutropenia observed in the palbociclib + letrozole arm of the overall safety population. Postmenopausal women (n = 165) with ER+, HER2-negative, advanced breast cancer who had not received any systemic treatment for their advanced disease were randomized 1:1 to receive either palbociclib in combination with letrozole or letrozole alone. Treatment continued until disease progression, unacceptable toxicity, consent withdrawal, or death. The primary endpoint was PFS. We now analyze the difference in PFS for the treatment populations by subgroups, including age, histological type, history of prior neoadjuvant/adjuvant systemic treatment, and sites of distant metastasis, using the Kaplan-Meier method. HR and 95 % CI are derived from a Cox proportional hazards regression model. A clinically meaningful improvement in median PFS and clinical benefit response (CBR) rate was seen with palbociclib + letrozole in every subgroup evaluated. Grade 3-4 neutropenia was the most common AE with palbociclib + letrozole in all subgroups. Analysis of the frequency of neutropenia by grade during the first six cycles of treatment showed that there was a downward trend in Grade 3-4 neutropenia

  5. Safety balance: Analysis of safety systems

    Delage, M.; Giroux, C.

    1990-12-01

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

  6. Status of the EU test blanket systems safety studies

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

    2015-01-01

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

  7. Status of the EU test blanket systems safety studies

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

    2015-10-15

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

  8. Further development and data basis for safety and accident analyses of nuclear front end and back end facilities and actualization and revision of calculation methods for nuclear safety analyses. Final report; Weiterentwicklung von Methoden und Datengrundlagen zu Sicherheits- und Stoerfallanalysen fuer Anlagen der nuklearen Ver- und Entsorgung sowie Aktualisierung und Ueberpruefung von Rechenmethoden zu nuklearen Sicherheitsanalysen. Abschlussbericht

    Kilger, Robert; Peters, Elisabeth; Sommer, Fabian; Moser, Eberhard-Franz; Kessen, Sven; Stuke, Maik

    2016-07-15

    This report briefly describes the activities carried out under the project 3613R03350 on the GRS ''Handbook on Accident Analysis for Nuclear Front and Back End Facilities'', and in detail the continuing work on the revision and updating of the GRS ''Handbook on Criticality'', which here focused on fissile systems with plutonium and {sup 233}U. The in previous projects started and ongoing literature study on innovative fuel concepts is continued. Also described are the review and qualification of computational methods by research and active benchmark participation, and the results of tracking the state of science and technology in the field of computational methods for criticality safety analysis. Special in-depth analyzes of selected criticality-relevant occurrences in the past are also documented.

  9. FOOD SAFETY IN CATERING INDUSTRY

    P. Cattaneo

    2009-03-01

    Full Text Available Catering industry plays a very important role in public health management, because about 30% of total daily meals are consumed in catering industry (restaurants, bar. In this work food safety was evaluated in 20 catering centres throughout microbiological analyses of different categories of meals. Results demonstrate that there was an important decrease of microbial contamination between 2006 and 2007, no pathogens were found in 217 samples examined: this was obtained by improving voluntary controls.

  10. Combination of cisplatin/S-1 in the treatment of patients with advanced gastric or gastroesophageal adenocarcinoma: Results of noninferiority and safety analyses compared with cisplatin/5-fluorouracil in the First-Line Advanced Gastric Cancer Study.

    Ajani, J A; Buyse, M; Lichinitser, M; Gorbunova, V; Bodoky, G; Douillard, J Y; Cascinu, S; Heinemann, V; Zaucha, R; Carrato, A; Ferry, D; Moiseyenko, V

    2013-11-01

    The aim of developing oral fluorouracil (5-FU) is to provide a more convenient administration route with similar efficacy and the best achievable tolerance. S-1, a novel oral fluoropyrimidine, was specifically designed to overcome the limitations of intravenous fluoropyrimidine therapies. A multicentre, randomised phase 3 trial was undertaken to compare S-1/cisplatin (CS) with infusional 5-FU/cisplatin (CF) in 1053 patients with untreated, advanced gastric/gastroesophageal adenocarcinoma. This report discusses a post-hoc noninferiority overall survival (OS) and safety analyses. Results (1029 treated; CS = 521/CF = 508) revealed OS in CS (8.6 months) was statistically noninferior to CF (7.9 months) [hazard ratio (HR) = 0.92 (two-sided 95% confidence interval (CI), 0.80-1.05)] for any margin equal to or greater than 1.05. Statistically significant safety advantages for the CS arm were observed [G3/4 neutropenia (CS, 18.6%; CF, 40.0%), febrile neutropenia (CS, 1.7%; CF, 6.9%), G3/4 stomatitis (CS, 1.3%; CF, 13.6%), diarrhoea (all grades: CS, 29.2%; CF, 38.4%) and renal adverse events (all grades: CS, 18.8%; CF, 33.5%)]. Hand-foot syndrome, infrequently reported, was mainly grade 1/2 in both arms. Treatment-related deaths were significantly lower in the CS arm than the CF arm (2.5% and 4.9%, respectively; Psafety profile and provides a new treatment option for patients with advanced gastric carcinoma. Copyright © 2013 Elsevier Ltd. All rights reserved.

  11. AEGIS methodology and a perspective from AEGIS methodology demonstrations

    Dove, F.H.

    1981-03-01

    Objectives of AEGIS (Assessment of Effectiveness of Geologic Isolation Systems) are to develop the capabilities needed to assess the post-closure safety of waste isolation in geologic formation; demonstrate these capabilities on reference sites; apply the assessment methodology to assist the NWTS program in site selection, waste package and repository design; and perform repository site analyses for the licensing needs of NWTS. This paper summarizes the AEGIS methodology, the experience gained from methodology demonstrations, and provides an overview in the following areas: estimation of the response of a repository to perturbing geologic and hydrologic events; estimation of the transport of radionuclides from a repository to man; and assessment of uncertainties

  12. Human factors in safety assessment. Safety culture assessment

    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

  13. Traceability of Software Safety Requirements in Legacy Safety Critical Systems

    Hill, Janice L.

    2007-01-01

    How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

  14. Demonstration and Dialogue: Mediation in Swedish Nuclear Waste Management

    Elam, Mark; Lidberg, Maria; Soneryd, Linda; Sundqvist, Goeran

    2009-01-01

    This report analyses mediation and mediators in Swedish nuclear waste management. Mediation is about establishing agreement and building common knowledge. It is argued that demonstrations and dialogue are the two prominent approaches to mediation in Swedish nuclear waste management. Mediation through demonstration is about showing, displaying, and pointing out a path to safe disposal for inspection. It implies a strict division between demonstrator and audience. Mediation through dialogue on the other hand, is about collective acknowledgements of uncertainty and suspensions of judgement creating room for broader discussion. In Sweden, it is the Swedish Nuclear Fuel and Waste Management Co. (SKB) that is tasked with finding a method and a site for the final disposal of the nation's nuclear waste. Two different legislative frameworks cover this process. In accordance with the Act on Nuclear Activities, SKB is required to demonstrate the safety of its planned nuclear waste management system to the government, while in respect of the Swedish Environmental Code, they are obliged to organize consultations with the public. How SKB combines these requirements is the main question under investigation in this report in relation to materials deriving from three empirical settings: 1) SKB's safety analyses, 2) SKB's public consultation activities and 3) the 'dialogue projects', initiated by other actors than SKB broadening the public arena for discussion. In conclusion, an attempt is made to characterise the long- term interplay of demonstration and dialogue in Swedish nuclear waste management

  15. Demonstration of the Safety and Feasibility of Robotically Assisted Percutaneous Coronary Intervention in Complex Coronary Lesions: Results of the CORA-PCI Study (Complex Robotically Assisted Percutaneous Coronary Intervention).

    Mahmud, Ehtisham; Naghi, Jesse; Ang, Lawrence; Harrison, Jonathan; Behnamfar, Omid; Pourdjabbar, Ali; Reeves, Ryan; Patel, Mitul

    2017-07-10

    in the robotic group (42:59 ± 26:14 min:s with R-PCI vs. 34:01 ± 17:14 min:s with M-PCI; p = 0.007), although clinical success remained similar (98.8% with R-PCI vs. 100% with M-PCI; p = 1.00). This study demonstrates the feasibility, safety, and high technical success of R-PCI for the treatment of complex coronary disease. Furthermore, comparable clinical outcomes, without an adverse effect on stent use or fluoroscopy time, were observed with R-PCI and M-PCI. Copyright © 2017 American College of Cardiology Foundation. Published by Elsevier Inc. All rights reserved.

  16. Independent assessment for new nuclear reactor safety

    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.

  17. Independent assessment for new nuclear reactor safety

    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)

  18. Advanced handbook for accident analyses of German nuclear power plants; Weiterentwicklung eines Handbuches fuer Stoerfallanalysen deutscher Kernkraftwerke

    Kerner, Alexander; Broecker, Annette; Hartung, Juergen; Mayer, Gerhard; Pallas Moner, Guim

    2014-09-15

    The advanced handbook of safety analyses (HSA) comprises a comprehensive electronic collection of knowledge for the compilation and conduction of safety analyses in the area of reactor, plant and containment behaviour as well as results of existing safety analyses (performed by GRS in the past) with characteristic specifications and further background information. In addition, know-how from the analysis software development and validation process is presented and relevant rules and regulations with regard to safety demonstration are provided. The HSA comprehensively covers the topic thermo-hydraulic safety analyses (except natural hazards, man-made hazards and malicious acts) for German pressurized and boiling water reactors for power and non-power operational states. In principle, the structure of the HSA-content represents the analytical approach utilized by safety analyses and applying the knowledge from safety analyses to technical support services. On the basis of a multilevel preparation of information to the topics ''compilation of safety analyses'', ''compilation of data bases'', ''assessment of safety analyses'', ''performed safety analyses'', ''rules and regulation'' and ''ATHLET-validation'' the HSA addresses users with different background, allowing them to enter the HSA at different levels. Moreover, the HSA serves as a reference book, which is designed future-oriented, freely configurable related to the content, completely integrated into the GRS internal portal and prepared to be used by a growing user group.

  19. Regulatory considerations for computational requirements for nuclear criticality safety

    Bidinger, G.H.

    1995-01-01

    As part of its safety mission, the U.S. Nuclear Regulatory Commission (NRC) approves the use of computational methods as part of the demonstration of nuclear criticality safety. While each NRC office has different criteria for accepting computational methods for nuclear criticality safety results, the Office of Nuclear Materials Safety and Safeguards (NMSS) approves the use of specific computational methods and methodologies for nuclear criticality safety analyses by specific companies (licensees or consultants). By contrast, the Office of Nuclear Reactor Regulation approves codes for general use. Historically, computational methods progressed from empirical methods to one-dimensional diffusion and discrete ordinates transport calculations and then to three-dimensional Monte Carlo transport calculations. With the advent of faster computational ability, three-dimensional diffusion and discrete ordinates transport calculations are gaining favor. With the proper user controls, NMSS has accepted any and all of these methods for demonstrations of nuclear criticality safety

  20. A concurrent diagnosis of microbiological food safety output and food safety management system performance: Cases from meat processing industries

    Luning, P.A.; Jacxsens, L.; Rovira, J.; Oses Gomez, S.; Uyttendaele, M.; Marcelis, W.J.

    2011-01-01

    Stakeholder requirements force companies to analyse their food safety management system (FSMS) performance to improve food safety. Performance is commonly analysed by checking compliance against preset requirements via audits/inspections, or actual food safety (FS) output is analysed by

  1. EUROSAFE Forum for nuclear safety. Towards Convergence of Technical Nuclear Safety Practices in Europe. Safety Improvements - Reasons, Strategies, Implementation

    Erven, Ulrich (ed.) [Gesellschaft fuer Anlagen- und Reaktorsicherheit, GRS mbH, Schwertnergasse 1, 50667 Koeln (Germany); Cherie, Jean-Bernard (ed.) [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Boeck, Benoit De (ed.) [Association Vincotte Nuclear, AVN, Rue Walcourt 148, 1070 Bruxelles (Belgium)

    2005-07-01

    The EUROSAFE Forum for Nuclear Safety is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE Web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety. The goal is to share experiences, to exchange technical and scientific opinions, and to conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum on 2005 focused on Safety Improvements, Reasons - Strategies - Implementation, from the point of view of the authorities, TSOs and industry. Latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe are presented. A high level of nuclear safety is a priority for the countries of Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining

  2. EUROSAFE Forum for nuclear safety. Towards Convergence of Technical Nuclear Safety Practices in Europe. Safety Improvements - Reasons, Strategies, Implementation

    Erven, Ulrich [Gesellschaft fuer Anlagen- und Reaktorsicherheit, GRS mbH, Schwertnergasse 1, 50667 Koeln (Germany); Cherie, Jean-Bernard [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Boeck, Benoit De [Association Vincotte Nuclear, AVN, Rue Walcourt 148, 1070 Bruxelles (Belgium)

    2005-07-01

    The EUROSAFE Forum for Nuclear Safety is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE Web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety. The goal is to share experiences, to exchange technical and scientific opinions, and to conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum on 2005 focused on Safety Improvements, Reasons - Strategies - Implementation, from the point of view of the authorities, TSOs and industry. Latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe are presented. A high level of nuclear safety is a priority for the countries of Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining

  3. Playground Safety

    ... Prevention Fall Prevention Playground Safety Poisoning Prevention Road Traffic Safety Sports Safety Get Email Updates To receive ... at the Consumer Product Safety Commission’s Playground Safety website . References U.S. Consumer Product Safety Commission. Injuries and ...

  4. Tofacitinib, an oral Janus kinase inhibitor, for the treatment of Latin American patients with rheumatoid arthritis: Pooled efficacy and safety analyses of Phase 3 and long-term extension studies.

    Radominski, Sebastião Cezar; Cardiel, Mario Humberto; Citera, Gustavo; Goecke, Annelise; Jaller, Juan Jose; Lomonte, Andrea Barranjard Vannucci; Miranda, Pedro; Velez, Patricia; Xibillé, Daniel; Kwok, Kenneth; Rojo, Ricardo; García, Erika Gabriela

    Tofacitinib is an oral Janus kinase inhibitor for the treatment of rheumatoid arthritis (RA). We assessed tofacitinib efficacy and safety in the Latin American (LA) subpopulation of global Phase 3 and long-term extension (LTE) studies. Data from LA patients with RA and inadequate response to disease-modifying antirheumatic drugs (DMARDs) were pooled across five Phase 3 studies. Phase 3 patients received tofacitinib 5 or 10mg twice daily (BID), adalimumab or placebo; patients in the single LTE study received tofacitinib 5 or 10mg BID; treatments were administered alone or with conventional synthetic DMARDs. Efficacy was reported up to 12 months (Phase 3) and 36 months (LTE) by American College of Rheumatology (ACR) 20/50/70 response rates, Disease Activity Score (DAS)28-4(erythrocyte sedimentation rate [ESR]) and Health Assessment Questionnaire-Disability Index (HAQ-DI). Incidence rates (IRs; patients with event/100 patient-years) of adverse events (AEs) of special interest were reported. The Phase 3 studies randomized 496 LA patients; the LTE study enrolled 756 LA patients from Phase 2 and Phase 3. In the Phase 3 studies, patients who received tofacitinib 5 and 10mg BID showed improvements vs placebo at Month 3 in ACR20 (68.9% and 75.7% vs 35.6%), ACR50 (45.8% and 49.7% vs 20.7%) and ACR70 (17.5% and 23.1% vs 6.9%) responses, mean change from baseline in HAQ-DI (-0.6 and -0.8 vs -0.3) and DAS28-4(ESR) score (-2.3 and -2.4 vs -1.4). The improvements were sustained up to Month 36 in the LTE study. In the Phase 3 studies, IRs with tofacitinib 5 and 10mg BID and placebo were 7.99, 6.57 and 9.84, respectively, for SAEs, and 3.87, 5.28 and 3.26 for discontinuation due to AEs. IRs of AEs of special interest in tofacitinib-treated LA patients were similar to the global population. In Phase 3 and LTE studies in LA patients with RA, tofacitinib demonstrated efficacy up to 36 months with a manageable safety profile up to 60 months, consistent with the overall tofacitinib

  5. Safety analysis and synthesis using fuzzy sets and evidential reasoning

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

    1995-01-01

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

  6. Impact of prior treatment status and reasons for discontinuation on the efficacy and safety of fingolimod: Subgroup analyses of the Fingolimod Research Evaluating Effects of Daily Oral Therapy in Multiple Sclerosis (FREEDOMS) study.

    Kremenchutzky, Marcelo; O'Connor, Paul; Hohlfeld, Reinhard; Zhang-Auberson, Lixin; von Rosenstiel, Philipp; Meng, Xiangyi; Grinspan, Augusto; Hashmonay, Ron; Kappos, Ludwig

    2014-05-01

    Fingolimod is a once-daily, oral sphingosine 1-phosphate receptor modulator approved for the treatment of relapsing multiple sclerosis. This post-hoc analysis of phase 3 FREEDOMS data assessed whether the effects of fingolimod are consistent among subgroups of patients defined by prior treatment history. Annualized relapse rate and safety profile of treatment with fingolimod 0.5mg, 1.25mg, or placebo once-daily for 24 months were analyzed in 1272 relapsing multiple sclerosis patients, by subgroups based on disease-modifying therapy history (treatment-naive; prior interferon-β or glatiramer acetate), reason for discontinuation of prior disease-modifying therapy (unsatisfactory therapeutic response or adverse events), and prior disease-modifying therapy duration. Both fingolimod doses significantly reduced annualized relapse rate in patients that received prior interferon-β or glatiramer acetate, discontinued prior disease-modifying therapy owing to unsatisfactory therapeutic effect, were treatment-naive, or had prior disease-modifying therapy duration of >1-3 years (P≤0.0301 for all comparisons vs placebo). Fingolimod 1.25mg resulted in greater reductions in annualized relapse rate in patients that discontinued prior disease-modifying therapy for adverse events or had prior disease-modifying therapy duration of ≤1 year or >3 years (P≤0.0194 vs placebo). Fingolimod demonstrated similar efficacy in relapsing multiple sclerosis patients regardless of prior treatment history. Clinicaltrials.gov identifier: NCT00289978. © 2013 The Authors. Published by Elsevier B.V. All rights reserved.

  7. NPP Temelin safety analysis reports and PSA status

    Mlady, O.

    1999-01-01

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

  8. Plutonium Finishing Plant safety evaluation report

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

  9. Patient safety: Safety culture and patient safety ethics

    Madsen, Marlene Dyrløv

    2006-01-01

    ,demonstrating significant, consistent and sometimes large differences in terms of safety culture factors across the units participating in the survey. Paper 5 is the results of a study of the relation between safety culture, occupational health andpatient safety using a safety culture questionnaire survey......Patient safety - the prevention of medical error and adverse events - and the initiative of developing safety cultures to assure patients from harm have become one of the central concerns in quality improvement in healthcare both nationally andinternationally. This subject raises numerous...... challenging issues of systemic, organisational, cultural and ethical relevance, which this dissertation seeks to address through the application of different disciplinary approaches. The main focus of researchis safety culture; through empirical and theoretical studies to comprehend the phenomenon, address...

  10. Applicability of ICRP principles for safety analysis of radioactive waste geological storage; Etude de l'applicabilite des principes de la CIPR a l'analyse de surete du stockage geologique des dechets radioactifs

    Lombard, J; Hubert, P; Pages, P

    1987-07-01

    Since the beginning of the eighties, the international organisations have established new recommendations for radioactive waste management. These recommendations are based on two principles. First is concerned with limitation of risks. It should be shown that the risk is smaller than the limit of acceptance. Practically only on risk criterion is foreseen. The principle demands, if a storage causes an event of individual risk (defined as a product of probability of occurrence and the probability of its causing severe health effects) is higher than 10 {sup -5} per year, this storage is not acceptable. The second principle deals with optimisation, demands that the level of protection related to the storage should be determined by a comparative process choosing the best compromise between the price of protection and the residual risk. These recommendations, especially the second one, differ from the safety analysis principles adopted presently in France and other countries. This study analyzes the advantages and potential inconveniences related to the introduction of the second principle. (author) [French] Depuis le debut des annees quatre vingt, de nouvelles recommandations ont ete formulees par les organismes internationaux (AIEA, OCDE, OIPR) en matiere de gestion des dechets radioactifs. Ces recommandations s'articulent autour de deux principes. Le premier, est celui de limitation des risques. II s'agit de demontrer que le risque est inferieur a un seuil d'acceptabilite. En pratique, un seul critere de risque est envisage. Ainsi le principe stipule, que si un stockage est a I'origine d'evenements conduisant a un risque individuel (defini comme le produit de a probabilite d'occurrence de l'evenement par Ia probabilite que cet evenement cause un effet sanitaire grave) total annuel superieur a 10{sup -5}/an, alors le risque lie a ce stockage est juge inacceptable. Le second, dit d'optimisation, stipule que le niveau de protection associe a un stockage (donc les

  11. ITER-FEAT safety

    Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

    2001-01-01

    Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

  12. Safety analysis report for packaging (onsite) steel drum

    McCormick, W.A.

    1998-01-01

    This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum

  13. Radiation safety

    Van Riessen, A.

    2002-01-01

    Full text: Experience has shown that modem, fully enclosed, XRF and XRD units are generally safe. This experience may lead to complacency and ultimately a lowering of standards which may lead to accidents. Maintaining awareness of radiation safety issues is thus an important role for all radiation safety officers. With the ongoing progress in technology, a greater number of radiation workers are more likely to use a range of instruments/techniques - eg portable XRF, neutron beam analysis, and synchrotron radiation analysis. The source for each of these types of analyses is different and necessitates an understanding of the associated dangers as well as use of specific radiation badges. The trend of 'suitcase science' is resulting in scientists receiving doses from a range of instruments and facilities with no coordinated approach to obtain an integrated dose reading for an individual. This aspect of radiation safety needs urgent attention. Within Australia a divide is springing up between those who work on Commonwealth property and those who work on State property. For example a university staff member may operate irradiating equipment on a University campus and then go to a CSIRO laboratory to operate similar equipment. While at the University State regulations apply and while at CSIRO Commonwealth regulations apply. Does this individual require two badges? Is there a need to obtain two licences? The application of two sets of regulations causes unnecessary confusion and increases the workload of radiation safety officers. Radiation safety officers need to introduce risk management strategies to ensure that both existing and new procedures result in risk minimisation. A component of this strategy includes ongoing education and revising of regulations. AXAA may choose to contribute to both of these activities as a service to its members as well as raising the level of radiation safety for all radiation workers. Copyright (2002) Australian X-ray Analytical

  14. Project CHRISTA. Feasibility study on the development of a safety demonstration methodology for a final repository for heat generating radioactive wastes in crystalline rock formations in Germany; Projekt CHRISTA. Machbarkeitsuntersuchung zur Entwicklung einer Sicherheits- und Nachweismethodik fuer ein Endlager fuer Waerme entwickelnde radioaktive Abfaelle im Kristallingestein in Deutschland

    Jobmann, Michael (ed.)

    2016-10-24

    In the frame of CHRISTA several options with different safety concepts for the final disposal of heat generating radioactive wastes were studied. The German safety requirements and the demonstration of the geological barrier integrity are based on an enclosure concept (ewG) that was developed primarily for salt and clay formations. The applicability of these requirements for crystalline host rocks had to be investigated. The enclosure functio0n is based on low hydraulic permeability of the host rock in combination with geotechnical barriers closing the access. With respect to the transferability of the Swedish/Finnish KBS-3 concept it has to be remarked, that the national standards in Sweden and Finland require the safety demonstration for 100.000 years (in Germany 1 million years). The Swedish/Finish container concept is based on a copper sheathed container with adjacent buffer; MOX fuel elements are not foreseen. The report concludes that the actual German safety concept based on geological barriers is to be preferred compared to technical barriers.

  15. Leadership style and patient safety: implications for nurse managers.

    Merrill, Katreena Collette

    2015-06-01

    The purpose of this study was to explore the relationship between nurse manager (NM) leadership style and safety climate. Nursing leaders are needed who will change the environment and increase patient safety. Hospital NMs are positioned to impact day-to-day operations. Therefore, it is essential to inform nurse executives regarding the impact of leadership style on patient safety. A descriptive correlational study was conducted in 41 nursing departments across 9 hospitals. The hospital unit safety climate survey and multifactorial leadership questionnaire were completed by 466 staff nurses. Bivariate and regression analyses were conducted to determine how well leadership style predicted safety climate. Transformational leadership style was demonstrated as a positive contributor to safety climate, whereas laissez-faire leadership style was shown to negatively contribute to unit socialization and a culture of blame. Nursing leaders must concentrate on developing transformational leadership skills while also diminishing negative leadership styles.

  16. Link between Research, Development and Demonstration (RD and D) and Stakeholder Confidence: the Specific Aspect of Long-term Safety. Proceedings of a Topical Session, Issy-les-Moulineaux, France 6-8 June 2007

    2008-01-01

    The Forum on Stakeholder Confidence (FSC) was created under a mandate from the OECD Nuclear Energy Agency's Radioactive Waste Management Committee (RWMC) to facilitate the sharing of international experience in addressing the societal dimension of radioactive waste management. It explores means of ensuring an effective dialogue with the public, and considers ways to strengthen confidence in decision-making processes. NEA member countries nominate participants. The FSC today includes representatives of national regulators, implementing agencies, policy makers, and R and D scientists from 16 OECD countries and two international organisations. Each has experience in and/or responsibility for stakeholder interaction. The Forum was launched in August 2000, in Paris, and convenes a series of alternating regular meetings and workshops. The FSC is expected to identify specific issues of interest on which stakeholders can learn from one another and provide a platform for discussing and exchanging those issues. This topical session's emphasis on long-term safety is meant to complement the Programme of Work of the RWMC in this area and to create a bridge between the FSC and the more technically oriented groups of the RWMC, e.g., the Integration Group for the Safety Case (IGSC) and the Regulators' Forum. These Proceedings are meant to provide a record that can be used by a wide spectrum of stakeholders and decision-makers. They include: a) An executive summary of the topical presentations, and of the discussion they sparked within the FSC; b) Texts and Power Point presentations provided by the speakers at the Topical Session

  17. Safety analysis procedures for PHWR

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

    2004-03-01

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

  18. Methodology of cost benefit analyses

    Patrik, M.; Babic, P.

    2000-10-01

    The report addresses financial aspects of proposed investments and other steps which are intended to contribute to nuclear safety. The aim is to provide introductory insight into the procedures and potential of cost-benefit analyses as a routine guide when making decisions on costly provisions as one of the tools to assess whether a particular provision is reasonable. The topic is applied to the nuclear power sector. (P.A.)

  19. Safety climate and culture: Integrating psychological and systems perspectives.

    Casey, Tristan; Griffin, Mark A; Flatau Harrison, Huw; Neal, Andrew

    2017-07-01

    Safety climate research has reached a mature stage of development, with a number of meta-analyses demonstrating the link between safety climate and safety outcomes. More recently, there has been interest from systems theorists in integrating the concept of safety culture and to a lesser extent, safety climate into systems-based models of organizational safety. Such models represent a theoretical and practical development of the safety climate concept by positioning climate as part of a dynamic work system in which perceptions of safety act to constrain and shape employee behavior. We propose safety climate and safety culture constitute part of the enabling capitals through which organizations build safety capability. We discuss how organizations can deploy different configurations of enabling capital to exert control over work systems and maintain safe and productive performance. We outline 4 key strategies through which organizations to reconcile the system control problems of promotion versus prevention, and stability versus flexibility. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  20. Safety margins in deterministic safety analysis

    Viktorov, A.

    2011-01-01

    The concept of safety margins has acquired certain prominence in the attempts to demonstrate quantitatively the level of the nuclear power plant safety by means of deterministic analysis, especially when considering impacts from plant ageing and discovery issues. A number of international or industry publications exist that discuss various applications and interpretations of safety margins. The objective of this presentation is to bring together and examine in some detail, from the regulatory point of view, the safety margins that relate to deterministic safety analysis. In this paper, definitions of various safety margins are presented and discussed along with the regulatory expectations for them. Interrelationships of analysis input and output parameters with corresponding limits are explored. It is shown that the overall safety margin is composed of several components each having different origins and potential uses; in particular, margins associated with analysis output parameters are contrasted with margins linked to the analysis input. While these are separate, it is possible to influence output margins through the analysis input, and analysis method. Preserving safety margins is tantamount to maintaining safety. At the same time, efficiency of operation requires optimization of safety margins taking into account various technical and regulatory considerations. For this, basic definitions and rules for safety margins must be first established. (author)

  1. Seismic fragility analyses

    Kostov, Marin

    2000-01-01

    In the last two decades there is increasing number of probabilistic seismic risk assessments performed. The basic ideas of the procedure for performing a Probabilistic Safety Analysis (PSA) of critical structures (NUREG/CR-2300, 1983) could be used also for normal industrial and residential buildings, dams or other structures. The general formulation of the risk assessment procedure applied in this investigation is presented in Franzini, et al., 1984. The probability of failure of a structure for an expected lifetime (for example 50 years) can be obtained from the annual frequency of failure, β E determined by the relation: β E ∫[d[β(x)]/dx]P(flx)dx. β(x) is the annual frequency of exceedance of load level x (for example, the variable x may be peak ground acceleration), P(fI x) is the conditional probability of structure failure at a given seismic load level x. The problem leads to the assessment of the seismic hazard β(x) and the fragility P(fl x). The seismic hazard curves are obtained by the probabilistic seismic hazard analysis. The fragility curves are obtained after the response of the structure is defined as probabilistic and its capacity and the associated uncertainties are assessed. Finally the fragility curves are combined with the seismic loading to estimate the frequency of failure for each critical scenario. The frequency of failure due to seismic event is presented by the scenario with the highest frequency. The tools usually applied for probabilistic safety analyses of critical structures could relatively easily be adopted to ordinary structures. The key problems are the seismic hazard definitions and the fragility analyses. The fragility could be derived either based on scaling procedures or on the base of generation. Both approaches have been presented in the paper. After the seismic risk (in terms of failure probability) is assessed there are several approaches for risk reduction. Generally the methods could be classified in two groups. The

  2. Biodiesel Mass Transit Demonstration

    2010-04-01

    The Biodiesel Mass Transit Demonstration report is intended for mass transit decision makers and fleet managers considering biodiesel use. This is the final report for the demonstration project implemented by the National Biodiesel Board under a gran...

  3. Authoring Effective Demonstrations

    Fu, Dan; Jensen, Randy; Salas, Eduardo; Rosen, Michael A; Ramachandran, Sowmya; Upshaw, Christin L; Hinkelman, Elizabeth; Lampton, Don

    2007-01-01

    ... or human role-players for each training event. We report our ongoing efforts to (1) research the nature and purpose of demonstration, articulating guidelines for effective demonstration within a training context, and (2...

  4. Comparing Demonstratives in Kwa

    This paper is a comparative study of demonstrative forms in three K wa languages, ... relative distance from the deictic centre, such as English this and that, here and there. ... Mostly, the referents of demonstratives are 'activated' or at least.

  5. Polarized Light Corridor Demonstrations.

    Davies, G. R.

    1990-01-01

    Eleven demonstrations of light polarization are presented. Each includes a brief description of the apparatus and the effect demonstrated. Illustrated are strain patterns, reflection, scattering, the Faraday Effect, interference, double refraction, the polarizing microscope, and optical activity. (CW)

  6. Safety study application guide

    1993-07-01

    Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

  7. Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

    2012-12-01

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

  8. Auto Safety

    ... Safe Videos for Educators Search English Español Auto Safety KidsHealth / For Parents / Auto Safety What's in this ... by teaching some basic rules. Importance of Child Safety Seats Using a child safety seat (car seat) ...

  9. Offsite demonstrations for MWLID technologies

    Williams, C.; Gruebel, R.

    1995-01-01

    The goal of the Offsite Demonstration Project for Mixed Waste Landfill Integrated Demonstration (MWLID)-developed environmental site characterization and remediation technologies is to facilitate the transfer, use, and commercialization of these technologies to the public and private sector. The meet this goal, the project identified environmental restoration needs of mixed waste and/or hazardous waste landfill owners (Native American, municipal, DOE, and DoD); documenting potential demonstration sites and the contaminants present at each site; assessing the environmental regulations that would effect demonstration activities; and evaluating site suitability for demonstrating MWLID technologies at the tribal and municipal sites identified. Eighteen landfill sites within a 40.2-km radius of Sandia National Laboratories are listed on the CERCLIS Site/Event Listing for the state of New Mexico. Seventeen are not located within DOE or DoD facilities and are potential offsite MWLID technology demonstration sites. Two of the seventeen CERCLIS sites, one on Native American land and one on municipal land, were evaluated and identified as potential candidates for off-site demonstrations of MWLID-developed technologies. Contaminants potentially present on site include chromium waste, household/commercial hazardous waste, volatile organic compounds, and petroleum products. MWLID characterization technologies applicable to these sites include Magnetometer Towed Array, Cross-borehole Electromagnetic Imaging, SitePlanner trademark/PLUME, Hybrid Directional Drilling, Seamist trademark/Vadose Zone Monitoring, Stripping Analyses, and x-ray Fluorescence Spectroscopy for Heavy Metals

  10. Strategy Guideline: Demonstration Home

    Savage, C.; Hunt, A.

    2012-12-01

    This guideline will provide a general overview of the different kinds of demonstration home projects, a basic understanding of the different roles and responsibilities involved in the successful completion of a demonstration home, and an introduction into some of the lessons learned from actual demonstration home projects. Also, this guideline will specifically look at the communication methods employed during demonstration home projects. And lastly, we will focus on how to best create a communication plan for including an energy efficient message in a demonstration home project and carry that message to successful completion.

  11. Strategy Guideline. Demonstration Home

    Hunt, A.; Savage, C.

    2012-12-01

    This guideline will provide a general overview of the different kinds of demonstration home projects, a basic understanding of the different roles and responsibilities involved in the successful completion of a demonstration home, and an introduction into some of the lessons learned from actual demonstration home projects. Also, this guideline will specifically look at the communication methods employed during demonstration home projects. And lastly, we will focus on how to best create a communication plan for including an energy efficient message in a demonstration home project and carry that message to successful completion.

  12. Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range; Sicherheitstechnische Analyse und Auslegungsaspekte von Abschirmungen gegen Teilchenstrahlung am Beispiel von Spallationsanlagen im Megawatt Bereich

    Hanslik, R.

    2006-08-15

    This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

  13. Safety KPIs - Monitoring of safety performance

    Andrej Lališ

    2014-09-01

    Full Text Available This paper aims to provide brief overview of aviation safety development focusing on modern trends represented by implementation of Safety Key Performance Indicators. Even though aviation is perceived as safe means of transport, it is still struggling with its complexity given by long-term growth and robustness which it has reached today. Thus nowadays safety issues are much more complex and harder to handle than ever before. We are more and more concerned about organizational factors and control mechanisms which have potential to further increase level of aviation safety. Within this paper we will not only introduce the concept of Key Performance Indicators in area of aviation safety as an efficient control mechanism, but also analyse available legislation and documentation. Finally we will propose complex set of indicators which could be applied to Czech Air Navigation Service Provider.

  14. Uncertainty Analyses and Strategy

    Kevin Coppersmith

    2001-01-01

    The DOE identified a variety of uncertainties, arising from different sources, during its assessment of the performance of a potential geologic repository at the Yucca Mountain site. In general, the number and detail of process models developed for the Yucca Mountain site, and the complex coupling among those models, make the direct incorporation of all uncertainties difficult. The DOE has addressed these issues in a number of ways using an approach to uncertainties that is focused on producing a defensible evaluation of the performance of a potential repository. The treatment of uncertainties oriented toward defensible assessments has led to analyses and models with so-called ''conservative'' assumptions and parameter bounds, where conservative implies lower performance than might be demonstrated with a more realistic representation. The varying maturity of the analyses and models, and uneven level of data availability, result in total system level analyses with a mix of realistic and conservative estimates (for both probabilistic representations and single values). That is, some inputs have realistically represented uncertainties, and others are conservatively estimated or bounded. However, this approach is consistent with the ''reasonable assurance'' approach to compliance demonstration, which was called for in the U.S. Nuclear Regulatory Commission's (NRC) proposed 10 CFR Part 63 regulation (64 FR 8640 [DIRS 101680]). A risk analysis that includes conservatism in the inputs will result in conservative risk estimates. Therefore, the approach taken for the Total System Performance Assessment for the Site Recommendation (TSPA-SR) provides a reasonable representation of processes and conservatism for purposes of site recommendation. However, mixing unknown degrees of conservatism in models and parameter representations reduces the transparency of the analysis and makes the development of coherent and consistent probability statements about projected repository

  15. Manufacturing Demonstration Facility (MDF)

    Federal Laboratory Consortium — The U.S. Department of Energy Manufacturing Demonstration Facility (MDF) at Oak Ridge National Laboratory (ORNL) provides a collaborative, shared infrastructure to...

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

    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.

  17. Waste Isolation Pilot Plant Safety Analysis Report

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  18. Waste Isolation Pilot Plant Safety Analysis Report

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  19. The N-terminal-truncated recombinant fibrin(ogen)olytic serine protease improves its functional property, demonstrates in vivo anticoagulant and plasma defibrinogenation activity as well as pre-clinical safety in rodent model.

    Bora, Bandana; Gogoi, Debananda; Tripathy, Debabrata; Kurkalang, Sillarine; Ramani, Sheetal; Chatterjee, Anupam; Mukherjee, Ashis K

    2018-05-01

    An N-terminal truncated fibrino(geno)lytic serine protease gene encoding a ~42kDa protein from Bacillus cereus strain AB01 was produced by error prone PCR, cloned into pET19b vector, and expressed in E5 coli BL21 DE3 cells. The deletion of 24 amino acid residues from N-terminal of wild-type Bacifrinase improves the catalytic activity of [Bacifrinase (ΔN24)]. The anticoagulant potency of [Bacifrinase (ΔN24)] was comparable to Nattokinase and Warfarin and results showed that its anticoagulant action is contributed by progressive defibrinogenation and antiplatelet activities. Nonetheless, at the tested concentration of 2.0μM [Bacifrinase (ΔN24)] did not show in vitro cytotoxicity or chromosomal aberrations on human embryonic kidney cells-293 (HEK-293) and human peripheral blood lymphocytes (HPBL) cells. [Bacifrinase (ΔN24)], at a dose of 2mg/kg, did not show toxicity, adverse pharmacological effects, tissue necrosis or hemorrhagic effect after 72h of its administration in Swiss albino mice. However, at the tested doses of 0.125 to 0.5mg/kg, it demonstrated significant in anticoagulant effect as well as defibrinogenation after 6h of administration in mice. We propose that [Bacifrinase (ΔN24)] may serve as prototype for the development of potent drug to prevent hyperfibrinogenemia related disorders. Copyright © 2018 Elsevier B.V. All rights reserved.

  20. Safety Regulation Implemented by Gosatomnadzor of Russia

    Gutsalov, A.T.; Bukrinsky, A.M.

    2001-01-01

    The principles and approaches used by Gosatomnadzor of Russia in establishing safety goals are described. The link between safety goals and safety culture is demonstrated. Information on nuclear regulatory activities in Russia is also presented

  1. Thermal hydraulic analyses of LVR-15 research reactor with IRT-M fuel

    Macek, J.

    1997-01-01

    The LVR-15 pool-type research reactor has been in operation at the Nuclear Research Institute at Rez since 1955. Following a number of reconstructions and redesigning, the current reactor power is 15 MW. Thermal hydraulic analyses to demonstrate that the core heat will be safely removed during operation as well as in accident situations were performed based on methodology which had been specifically developed for the LVR-15 research reactor. This methodology was applied to stationary thermal hydraulic computations, as well as to transients, particularly with reactivity failure and loss of circulation pumps emergencies. The applied methodology and the core configuration as used in the Safety Report are described. The initial and boundary conditions are then considered and the summary of the calculated failures with regard to the defined safety limits is presented. The results of the core configuration analyses are also discussed with respect to meeting the safety limits and to the applicability of the methodology to this purpose

  2. Integrated therapy safety management system.

    Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

    2013-09-01

    The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an 'integrated therapy safety management' is drafted. This concept could serve as a basis to improve resilience. The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for 'integrated therapy safety management'. The concept is applied by way of example for the 'medication process' to demonstrate its practical implementation. The 'integrated therapy safety management' is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of 'bridge managers'. 'Bridge managers' anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the 'bridge managers' and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. © 2013 The Authors. British Journal of Clinical Pharmacology © 2013 The British Pharmacological Society.

  3. Integrated therapy safety management system

    Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

    2013-01-01

    Aims The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an ‘integrated therapy safety management’ is drafted. This concept could serve as a basis to improve resilience. Methods The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for ‘integrated therapy safety management’. The concept is applied by way of example for the ‘medication process’ to demonstrate its practical implementation. Results The ‘integrated therapy safety management’ is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of ‘bridge managers’. ‘Bridge managers’ anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the ‘bridge managers’ and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. Conclusions The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. PMID:24007448

  4. Leadership for safety: industrial experience.

    Flin, R; Yule, S

    2004-12-01

    The importance of leadership for effective safety management has been the focus of research attention in industry for a number of years, especially in energy and manufacturing sectors. In contrast, very little research into leadership and safety has been carried out in medical settings. A selective review of the industrial safety literature for leadership research with possible application in health care was undertaken. Emerging findings show the importance of participative, transformational styles for safety performance at all levels of management. Transactional styles with attention to monitoring and reinforcement of workers' safety behaviours have been shown to be effective at the supervisory level. Middle managers need to be involved in safety and foster open communication, while ensuring compliance with safety systems. They should allow supervisors a degree of autonomy for safety initiatives. Senior managers have a prime influence on the organisation's safety culture. They need to continuously demonstrate a visible commitment to safety, best indicated by the time they devote to safety matters.

  5. Nuclear safety

    1991-02-01

    This book reviews the accomplishments, operations, and problems faced by the defense Nuclear Facilities Safety Board. Specifically, it discusses the recommendations that the Safety Board made to improve safety and health conditions at the Department of Energy's defense nuclear facilities, problems the Safety Board has encountered in hiring technical staff, and management problems that could affect the Safety Board's independence and credibility

  6. Innovative technology demonstration

    Anderson, D.B.; Luttrell, S.P.; Hartley, J.N.; Hinchee, R.

    1992-04-01

    The Innovative Technology Demonstration (ITD) program at Tinker Air Force Base (TAFB), Oklahoma City, Oklahoma, will demonstrate the overall utility and effectiveness of innovative technologies for site characterization, monitoring, and remediation of selected contaminated test sites. The current demonstration test sites include a CERCLA site on the NPL list, located under a building (Building 3001) that houses a large active industrial complex used for rebuilding military aircraft, and a site beneath and surrounding an abandoned underground tank vault used for storage of jet fuels and solvents. The site under Building 3001 (the NW Test Site) is contaminated with TCE and Cr +6 ; the site with the fuel storage vault (the SW Tanks Site) is contaminated with fuels, BTEX and TCE. These sites and others have been identified for cleanup under the Air Force's Installation Restoration Program (IRP). This document describes the demonstrations that have been conducted or are planned for the TAFB

  7. Laser Communications Relay Demonstration

    National Aeronautics and Space Administration — LCRD is a minimum two year flight demonstration in geosynchronous Earth orbit to advance optical communications technology toward infusion into Deep Space and Near...

  8. Reactor safety research and safety technology. Pt. 2

    Theenhaus, R.; Wolters, J.

    1987-01-01

    The state of HTR safety research work reached permits a comprehensive and reliable answer to be given to questions which have been raised by the reactor accident at Chernobyl, regarding HTR safety. Together with the probability safety analyses, the way to a safety concept suitable for an HTR is cleared; instructions are given for design optimisation with regard to safety technique and economy. The consequences of a graphite fire, the neutron physics design and the consequenes of the lack of a safety containment are briefly described. (DG) [de

  9. Education Payload Operation - Demonstrations

    Keil, Matthew

    2009-01-01

    Education Payload Operation - Demonstrations (EPO-Demos) are recorded video education demonstrations performed on the International Space Station (ISS) by crewmembers using hardware already onboard the ISS. EPO-Demos are videotaped, edited, and used to enhance existing NASA education resources and programs for educators and students in grades K-12. EPO-Demos are designed to support the NASA mission to inspire the next generation of explorers.

  10. FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS

    GARVIN, L. J.; JENSEN, M. A.

    2004-04-13

    This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

  11. Buried Waste Integrated Demonstration

    1994-03-01

    The Buried Waste Integrated Demonstration (BWID) supports the applied research, development, demonstration, and evaluation of a suite of advanced technologies that offer promising solutions to the problems associated with the remediation of buried waste. BWID addresses the difficult remediation problems associated with DOE complex-wide buried waste, particularly transuranic (TRU) contaminated buried waste. BWID has implemented a systems approach to the development and demonstration of technologies that will characterize, retrieve, treat, and dispose of DOE buried wastes. This approach encompasses the entire remediation process from characterization to post-monitoring. The development and demonstration of the technology is predicated on how a technology fits into the total remediation process. To address all of these technological issues, BWID has enlisted scientific expertise of individuals and groups from within the DOE Complex, as well as experts from universities and private industry. The BWID mission is to support development and demonstration of a suite of technologies that, when integrated with commercially-available technologies, forms a comprehensive, remediation system for the effective and efficient remediation of buried waste throughout the DOE Complex. BWID will evaluate and validate demonstrated technologies and transfer this information and equipment to private industry to support the Office of Environmental Restoration (ER), Office of Waste Management (WM), and Office of Facility Transition (FT) remediation planning and implementation activities

  12. General aviation crash safety program at Langley Research Center

    Thomson, R. G.

    1976-01-01

    The purpose of the crash safety program is to support development of the technology to define and demonstrate new structural concepts for improved crash safety and occupant survivability in general aviation aircraft. The program involves three basic areas of research: full-scale crash simulation testing, nonlinear structural analyses necessary to predict failure modes and collapse mechanisms of the vehicle, and evaluation of energy absorption concepts for specific component design. Both analytical and experimental methods are being used to develop expertise in these areas. Analyses include both simplified procedures for estimating energy absorption capabilities and more complex computer programs for analysis of general airframe response. Full-scale tests of typical structures as well as tests on structural components are being used to verify the analyses and to demonstrate improved design concepts.

  13. Drug Safety

    ... over-the-counter drug. The FDA evaluates the safety of a drug by looking at Side effects ... clinical trials The FDA also monitors a drug's safety after approval. For you, drug safety means buying ...

  14. Nuclear safety

    Tarride, Bruno

    2015-10-01

    The author proposes an overview of methods and concepts used in the nuclear industry, at the design level as well as at the exploitation level, to ensure an acceptable safety level, notably in the case of nuclear reactors. He first addresses the general objectives of nuclear safety and the notion of acceptable risk: definition and organisation of nuclear safety (relationships between safety authorities and operators), notion of acceptable risk, deterministic safety approach and main safety principles (safety functions and confinement barriers, concept of defence in depth). Then, the author addresses the safety approach at the design level: studies of operational situations, studies of internal and external aggressions, safety report, design principles for important-for-safety systems (failure criterion, redundancy, failure prevention, safety classification). The next part addresses safety during exploitation and general exploitation rules: definition of the operation domain and of its limits, periodic controls and tests, management in case of incidents, accidents or aggressions

  15. Safety culture

    Keen, L.J.

    2003-01-01

    Safety culture has become a topic of increasing interest for industry and regulators as issues are raised on safety problems around the world. The keys to safety culture are organizational effectiveness, effective communications, organizational learning, and a culture that encourages the identification and resolution of safety issues. The necessity of a strong safety culture places an onus on all of us to continually question whether the safety measures already in place are sufficient, and are being applied. (author)

  16. Learning From Demonstration?

    Koch, Christian; Bertelsen, Niels Haldor

    2014-01-01

    Demonstration projects are often used in the building sector to provide a basis for using new processes and/or products. The climate change agenda implies that construction is not only required to deliver value for the customer, cost reductions and efficiency but also sustainable buildings....... This paper reports on an early demonstration project, the Building of a passive house dormitory in the Central Region of Denmark in 2006-2009. The project was supposed to deliver value, lean design, prefabrication, quality in sustainability, certification according to German standards for passive houses......, and micro combined heat and power using hydrogen. Using sociological and business economic theories of innovation, the paper discusses how early movers of innovation tend to obtain only partial success when demonstrating their products and often feel obstructed by minor details. The empirical work...

  17. Solar renovation demonstration projects

    Bruun Joergensen, O [ed.

    1998-10-01

    In the framework of the IEA SHC Programme, a Task on building renovation was initiated, `Task 20, Solar Energy in Building Renovation`. In a part of the task, Subtask C `Design of Solar Renovation Projects`, different solar renovation demonstration projects were developed. The objective of Subtask C was to demonstrate the application of advanced solar renovation concepts on real buildings. This report documents 16 different solar renovation demonstration projects including the design processes of the projects. The projects include the renovation of houses, schools, laboratories, and factories. Several solar techniques were used: building integrated solar collectors, glazed balconies, ventilated solar walls, transparent insulation, second skin facades, daylight elements and photovoltaic systems. These techniques are used in several simple as well as more complex system designs. (au)

  18. Biodenitrification demonstration test report

    Benear, A.K.; Murray, S.J.; Lahoda, E.J.; Leslie, J.W.; Patton, J.B.; Menako, C.R.

    1987-08-01

    A two-column biodenitrification (BDN) facility was constructed at the Feed Materials Production Center (FMPC) in 1985 and 1986 to test the feasibility of biological treatment for industrial nitrate-bearing waste water generated at FMPC. This demonstration facility comprises one-half of the proposed four-column production facility. A demonstration test was conducted over a four month period in 1987. The results indicate the proposed BDN production facility can process FMPC industrial wastewater in a continuous manner while maintaining an effluent that will consistently meet the proposed NPDES limits for combined nitrate nitrogen (NO 3 -N) and nitrite nitrogen (NO 2 -N). The proposed NPDES limits are 62 kg/day average and 124 kg/day maximum. These limits were proportioned to determine that the two-column demonstration facility should meet the limits of 31 kg/day average and 62 kg/day maximum

  19. Distributed picture compilation demonstration

    Alexander, Richard; Anderson, John; Leal, Jeff; Mullin, David; Nicholson, David; Watson, Graham

    2004-08-01

    A physical demonstration of distributed surveillance and tracking is described. The demonstration environment is an outdoor car park overlooked by a system of four rooftop cameras. The cameras extract moving objects from the scene, and these objects are tracked in a decentralized way, over a real communication network, using the information form of the standard Kalman filter. Each node therefore has timely access to the complete global picture and because there is no single point of failure in the system, it is robust. The demonstration system and its main components are described here, with an emphasis on some of the lessons we have learned as a result of applying a corpus of distributed data fusion theory and algorithms in practice. Initial results are presented and future plans to scale up the network are also outlined.

  20. Photovoltaic demonstration projects

    Gillett, W B; Hacker, R J; Kaut, W [eds.

    1991-01-01

    This book, the proceedings of the fourth PV-Contractors' Meeting organized by the Commission of the European Communities, Directorate-General for Energy, held at Brussels on 21 and 22 November 1989, provides an overview of the photovoltaic demonstration projects which have been supported in the framework of the Energy Demonstration Program since 1983. It includes reports by each of the contractors who submitted proposals in 1983, 1984, 1985 and 1986, describing progress with their projects. Summaries of the discussions held at the meeting, which included contractors whose projects were submitted in 1987, are also presented. The different technologies which are being demonstrated concern the modules, the cabling of the array, structure design, storage strategy and power conditioning. The various applications include desalination, communications, dairy farms, water pumping, and warning systems. Papers have been processed separately for inclusion on the data base.

  1. Electric vehicle demonstration

    Ouellet, M. [National Centre for Advanced Transportation, Saint-Jerome, PQ (Canada)

    2010-07-01

    The desirable characteristics of Canadian projects that demonstrate vehicle use in real-world operation and the appropriate mechanism to collect and disseminate the monitoring data were discussed in this presentation. The scope of the project was on passenger cars and light duty trucks operating in plug-in electric vehicle (PHEV) or battery electric vehicle modes. The presentation also discussed the funding, stakeholders involved, Canadian travel pattern analysis, regulatory framework, current and recent electric vehicle demonstration projects, and project guidelines. It was concluded that some demonstration project activities may have been duplicated as communication between the proponents was insufficient. It was recommended that data monitoring using automatic data logging with minimum reliance on logbooks and other user entry should be emphasized. figs.

  2. Innovative technology demonstrations

    Anderson, D.B.; Luttrell, S.P.; Hartley, J.N.

    1992-08-01

    Environmental Management Operations (EMO) is conducting an Innovative Technology Demonstration Program for Tinker Air Force Base (TAFB). Several innovative technologies are being demonstrated to address specific problems associated with remediating two contaminated test sites at the base. Cone penetrometer testing (CPT) is a form of testing that can rapidly characterize a site. This technology was selected to evaluate its applicability in the tight clay soils and consolidated sandstone sediments found at TAFB. Directionally drilled horizontal wells was selected as a method that may be effective in accessing contamination beneath Building 3001 without disrupting the mission of the building, and in enhancing the extraction of contamination both in ground water and in soil. A soil gas extraction (SGE) demonstration, also known as soil vapor extraction, will evaluate the effectiveness of SGE in remediating fuels and TCE contamination contained in the tight clay soil formations surrounding the abandoned underground fuel storage vault located at the SW Tanks Site. In situ sensors have recently received much acclaim as a technology that can be effective in remediating hazardous waste sites. Sensors can be useful for determining real-time, in situ contaminant concentrations during the remediation process for performance monitoring and in providing feedback for controlling the remediation process. Following the SGE demonstration, the SGE system and SW Tanks test site will be modified to demonstrate bioremediation as an effective means of degrading the remaining contaminants in situ. The bioremediation demonstration will evaluate a bioventing process in which the naturally occurring consortium of soil bacteria will be stimulated to aerobically degrade soil contaminants, including fuel and TCE, in situ

  3. Innovative technology demonstrations

    Anderson, D.B.; Hartley, J.N.; Luttrell, S.P.

    1992-04-01

    Currently, several innovative technologies are being demonstrated at Tinker Air Force Base (TAFB) to address specific problems associated with remediating two contaminated test sites at the base. Cone penetrometer testing (CPT) is a form of testing that can rapidly characterize a site. This technology was selected to evaluate its applicability in the tight clay soils and consolidated sandstone sediments found at TAFB. Directionally drilled horizontal wells have been successfully installed at the US Department of Energy's (DOE) Savannah River Site to test new methods of in situ remediation of soils and ground water. This emerging technology was selected as a method that may be effective in accessing contamination beneath Building 3001 without disrupting the mission of the building, and in enhancing the extraction of contamination both in ground water and in soil. A soil gas extraction (SGE) demonstration, also known as soil vapor extraction, will evaluate the effectiveness of SGE in remediating fuels and TCE contamination contained in the tight clay soil formations surrounding the abandoned underground fuel storage vault located at the SW Tanks Site. In situ sensors have recently received much acclaim as a technology that can be effective in remediating hazardous waste sites. Sensors can be useful for determining real-time, in situ contaminant concentrations during the remediation process for performance monitoring and in providing feedback for controlling the remediation process. A demonstration of two in situ sensor systems capable of providing real-time data on contamination levels will be conducted and evaluated concurrently with the SGE demonstration activities. Following the SGE demonstration, the SGE system and SW Tanks test site will be modified to demonstrate bioremediation as an effective means of degrading the remaining contaminants in situ

  4. SRP reactor safety evolution

    Rankin, D.B.

    1984-01-01

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  5. Software Safety Risk in Legacy Safety-Critical Computer Systems

    Hill, Janice L.; Baggs, Rhoda

    2007-01-01

    Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

  6. Gigashot Optical Laser Demonstrator

    Deri, R. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-13

    The Gigashot Optical Laser Demonstrator (GOLD) project has demonstrated a novel optical amplifier for high energy pulsed lasers operating at high repetition rates. The amplifier stores enough pump energy to support >10 J of laser output, and employs conduction cooling for thermal management to avoid the need for expensive and bulky high-pressure helium subsystems. A prototype amplifier was fabricated, pumped with diode light at 885 nm, and characterized. Experimental results show that the amplifier provides sufficient small-signal gain and sufficiently low wavefront and birefringence impairments to prove useful in laser systems, at repetition rates up to 60 Hz.

  7. Photovoltaic demonstration projects 2

    Gillett, W B; Hacker, R J [Halcrow (William) and Partners, Swindon (UK); Kaut, W [eds.

    1989-01-01

    This book, the proceedings of the third Photovoltaic Contractors' Meeting organised by the Commission of the European Communities, Directorate-General for Energy provides an overview of the photovoltaic demonstration projects which have been supported by the Energy Directorate of the Commission of the European Communities since 1983. It includes reports by each of the contractors who submitted proposals in 1983, 1984 and 1985, describing progress with their projects. The different technologies which are being demonstrated concern the modules, the cabling of the array, structure design, storage strategy and power conditioning. The various applications include powering of houses, villages, recreation centres, water desalination, communications, dairy farms, water pumping and warning systems. (author).

  8. Safety case plan 2008

    2008-07-01

    Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy) Posiva is preparing to submit the construction license application for a spent fuel repository by the end of the year 2012. The long-term safety section supporting the license application is based on a safety case, which, according to the internationally adopted definition, is a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. In 2005, Posiva presented a plan to prepare such a safety case. The present report provides a revised plan of the safety case contents mentioned above. The update of the safety case plan takes into account the recommendations made by the Radiation and Nuclear Safety Authority (STUK) about improving the focus and further developing the plan. Accordingly, particular attention is given to the quality management of the safety case work, the management of uncertainties and the scenario methodology. The quality management is based on the ISO 9001:2000 standard process thinking enhanced with special features arising from STUK's YVL Guides. The safety case production process is divided into four main sub-processes. The conceptualisation and methodology sub-process defines the framework for the assessment. The critical data handling and modelling sub-process links Posiva's main technical and scientific activities to the production of the safety case. The assessment sub-process analyses the consequences of the evolution of the disposal system in various scenarios, classified either as part of the expected evolution or as disruptive scenarios. The compliance and confidence sub-process is responsible for final evaluation of compliance of the assessment results with the regulatory criteria and the overall confidence in the safety case. As in the previous safety case plan, the safety case will be based on several reports, but

  9. LOFT integral test system final safety analysis report

    1974-03-01

    Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)

  10. Removing unreasonable conservatisms in DOE safety analysis

    BISHOP, G.E.

    1999-01-01

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

  11. Inseparable Phone Books Demonstration

    Balta, Nuri; Çetin, Ali

    2017-01-01

    This study is aimed at first introducing a well-known discrepant event; inseparable phone books and second, turning it into an experiment for high school or middle school students. This discrepant event could be used especially to indicate how friction force can be effective in producing an unexpected result. Demonstration, discussion, explanation…

  12. PHARUS ASAR demonstrator

    Smith, A.J.E.; Bree, R.J.P. van; Calkoen, C.J.; Dekker, R.J.; Otten, M.P.G.; Rossum, W.L. van

    2001-01-01

    PHARUS is a polarimetric phased array C-band Synthetic Aperture Radar (SAR), designed and built for airborne use. Advanced SAR (ASAR) data in image and alternating polarization mode have been simulated with PHARUS to demonstrate the use of Envisat for a number of typical SAR applications that are

  13. Demonstrating the Gas Laws.

    Holko, David A.

    1982-01-01

    Presents a complete computer program demonstrating the relationship between volume/pressure for Boyle's Law, volume/temperature for Charles' Law, and volume/moles of gas for Avagadro's Law. The programing reinforces students' application of gas laws and equates a simulated moving piston to theoretical values derived using the ideal gas law.…

  14. Astronomy LITE Demonstrations

    Brecher, Kenneth

    2006-12-01

    Project LITE (Light Inquiry Through Experiments) is a materials, software, and curriculum development project. It focuses on light, optics, color and visual perception. According to two recent surveys of college astronomy faculty members, these are among the topics most often included in the large introductory astronomy courses. The project has aimed largely at the design and implementation of hands-on experiences for students. However, it has also included the development of lecture demonstrations that employ novel light sources and materials. In this presentation, we will show some of our new lecture demonstrations concerning geometrical and physical optics, fluorescence, phosphorescence and polarization. We have developed over 200 Flash and Java applets that can be used either by teachers in lecture settings or by students at home. They are all posted on the web at http://lite.bu.edu. For either purpose they can be downloaded directly to the user's computer or run off line. In lecture demonstrations, some of these applets can be used to control the light emitted by video projectors to produce physical effects in materials (e.g. fluorescence). Other applets can be used, for example, to demonstrate that the human percept of color does not have a simple relationship with the physical frequency of the stimulating source of light. Project LITE is supported by Grant #DUE-0125992 from the NSF Division of Undergraduate Education.

  15. A Magnetic Circuit Demonstration.

    Vanderkooy, John; Lowe, June

    1995-01-01

    Presents a demonstration designed to illustrate Faraday's, Ampere's, and Lenz's laws and to reinforce the concepts through the analysis of a two-loop magnetic circuit. Can be made dramatic and challenging for sophisticated students but is suitable for an introductory course in electricity and magnetism. (JRH)

  16. Safety Teams: An Approach to Engage Students in Laboratory Safety

    Alaimo, Peter J.; Langenhan, Joseph M.; Tanner, Martha J.; Ferrenberg, Scott M.

    2010-01-01

    We developed and implemented a yearlong safety program into our organic chemistry lab courses that aims to enhance student attitudes toward safety and to ensure students learn to recognize, demonstrate, and assess safe laboratory practices. This active, collaborative program involves the use of student "safety teams" and includes…

  17. Deep Borehole Disposal Safety Analysis.

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

    2016-10-01

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

  18. Hospital safety culture in Taiwan: a nationwide survey using Chinese version Safety Attitude Questionnaire.

    Lee, Wui-Chiang; Wung, Hwei-Ying; Liao, Hsun-Hsiang; Lo, Chien-Ming; Chang, Fei-Ling; Wang, Pa-Chun; Fan, Angela; Chen, Hsin-Hsin; Yang, Han-Chuan; Hou, Sheng-Mou

    2010-08-10

    Safety activities have been initiated at many hospitals in Taiwan, but little is known about the safety culture at these hospitals. The aims of this study were to verify a safety culture survey instrument in Chinese and to assess hospital safety culture in Taiwan. The Taiwan Patient Safety Culture Survey was conducted in 2008, using the adapted Safety Attitude Questionnaire in Chinese (SAQ-C). Hospitals and their healthcare workers participated in the survey on a voluntary basis. The psychometric properties of the five SAQ-C dimensions were examined, including teamwork climate, safety climate, job satisfaction, perception of management, and working conditions. Additional safety measures were asked to assess healthcare workers' attitudes toward their collaboration with nurses, physicians, and pharmacists, respectively, and perceptions of hospitals' encouragement of safety reporting, safety training, and delivery delays due to communication breakdowns in clinical areas. The associations between the respondents' attitudes to each SAQ-C dimension and safety measures were analyzed by generalized estimating equations, adjusting for the clustering effects at hospital levels. A total of 45,242 valid questionnaires were returned from 200 hospitals with a mean response rate of 69.4%. The Cronbach's alpha was 0.792 for teamwork climate, 0.816 for safety climate, 0.912 for job satisfaction, 0.874 for perception of management, and 0.785 for working conditions. Confirmatory factor analyses demonstrated a good model fit for each dimension and the entire construct. The percentage of hospital healthcare workers holding positive attitude was 48.9% for teamwork climate, 45.2% for perception of management, 42.1% for job satisfaction, 37.2% for safety climate, and 31.8% for working conditions. There were wide variations in the range of SAQ-C scores in each dimension among hospitals. Compared to those without positive attitudes, healthcare workers with positive attitudes to each SAQ

  19. Multicenter analysis of tolerance and clinical safety of the extracellular MR contrast agent gadobenate dimeglumine (MultiHance {sup registered}); Multizentrische Analyse der Vertraeglichkeit und klinischen Sicherheit des extrazellulaeren MR-Kontrastmittels Gadobenat-Dimeglumin (MultiHance {sup registered})

    Herborn, Christoph U.; Jaeger-Booth, I.; Goyen, M. [Universitaetsklinikum Hamburg-Eppendorf, Medizinisches PraeventionsCentrum Hamburg (MPCH) (Germany); Lodemann, K.P. [R und D, Altana Pharma (Germany); Spinazzi, A. [Marketing, BRACCO Ltd (Italy)

    2009-07-15

    Purpose: Retrospective analysis of the occurrence of adverse events and the diagnostic efficacy of a paramagnetic contrast agent with weak intermittent protein binding and high relaxivity. Materials end methods: Postmarketing surveillance studies for gadobenate dimeglumine (MultiHance, BRACCO Altana Pharma, Constance) were conducted in Germany between 1998 and 2006 and then retrospectively analyzed. Demographic data, relevant comorbidities, and allergies were recorded. The safety and tolerability of MultiHance were logged on a standardized data sheet. Results: A total of 38568 patients were included in the study. 829 patients (2.1%) had a known intolerance against contrast media. The examined regions included the central nervous system, the liver, and the vascular bed. The injection rate with automated injectors (n = 10456) varied between 1.0 und 3.0 ml/sec in 86.5% of patients. Adverse events totaled 1.2%. 11 patients (0.03%) experienced serious adverse events. The most frequent findings were nausea, vomiting and a feeling of warmth. Conclusion: MultiHance is a safe and very well tolerated contrast agent for magnetic resonance imaging (MRI) with a profile and frequency of adverse events similar to other extracellular MR contrast materials. (orig.)

  20. Demonstration of innovative techniques for work zone safety data analysis

    2009-07-15

    Based upon the results of the simulator data analysis, additional future research can be : identified to validate the driving simulator in terms of similarities with Ohio work zones. For : instance, the speeds observed in the simulator were greater f...

  1. Demonstration study on shielding safety analysis code. 7

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    2000-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) To improve the detection sensitivity of pulse neutron measurement, two neutron detectors and some electronic circuits are added to the system constructed last year. (2) To estimate the neutron dose at the distant point from the facility instead of the commercialized rem-counter, a {sup 3}He detector with paraffin moderator is equipped to the system. (3) Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility was measured in the distance up to 300 m. The results show that the time structure of pulsed neutrons almost disappears at the further points than 150 m. (4) In the distance from 90 m to 300 m ordinal total counting method without gate pulse are applied to detect the neutrons. (5) The experimental results of space dependency up to 300 m is fitted fairly well by the Gui's response function. (author)

  2. Demonstration study on shielding safety analysis code (VI)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  3. Demonstration study on shielding safety analysis code (8)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

    2001-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

  4. Adaptive Seat Energy Absorbers for Enhanced Crash Safety: Technology Demonstration

    2016-08-01

    is no pressing need for MREA refinement. However, should an increase in MREA yield force be desired, the project team explored 2 simple refinements...team developed a servo motor controller and data acquisition program using dSPACE real-time system. From the preliminary test, the preloaded Terfenol...Technology; 1940. b Oberg E, editor. Machinery’s handbook: eighteenth edition. Norwalk (CT): Industrial Press ; 1968. c Quayle JP, editor. Kempe’s

  5. Remote monitoring demonstration

    Caskey, Susan; Olsen, John

    2006-01-01

    The recently upgraded remote monitoring system at the Joyo Experimental Reactor uses a DCM-14 camera module and GEMINI software. The final data is compatible both with the IAEA-approved GARS review software and the ALIS software that was used for this demonstration. Features of the remote monitoring upgrade emphasized compatibility with IAEA practice. This presentation gives particular attention to the selection process for meeting network security considerations at the O'arai site. The Joyo system is different from the NNCA's ACPF system, in that it emphasizes use of IAEA standard camera technology and data acquisition and transmission software. In the demonstration itself, a temporary virtual private network (VPN) between the meeting room and the server at Sandia in Albuquerque allowed attendees to observe data stored from routine transmissions from the Joyo Fresh Fuel Storage to Sandia. Image files from a fuel movement earlier in the month showed Joyo workers and IAEA inspectors carrying out a transfer. (author)

  6. Commercial incineration demonstration

    Borduin, L.C.; Neuls, A.S.

    1981-01-01

    Low-level radioactive wastes (LLW) generated by nuclear utilities presently are shipped to commercial burial grounds for disposal. Substantially increasing shipping and disposal charges have sparked renewed industry interest in incineration and other advanced volume reduction techniques as potential cost-saving measures. Repeated inquiries from industry sources regarding LLW applicability of the Los Alamos controlled-air incineration (CAI) design led DOE to initiate this commercial demonstration program in FY-1980. The selected program approach to achieving CAI demonstration at a utility site is a DOE sponsored joint effort involving Los Alamos, a nuclear utility, and a liaison subcontractor. Required development tasks and responsibilities of the particpants are described. Target date for project completion is the end of FY-1985

  7. Photovoltaic demonstration projects

    Kaut, W [Commission of the European Communities, Brussels (Belgium); Gillett, W B; Hacker, R J [Halcrow Gilbert Associates Ltd., Swindon (GB)

    1992-12-31

    This publication, comprising the proceedings of the fifth contractor`s meeting organized by the Commission of the European Communities, Directorate-General for Energy, provides an overview of the photovoltaic demonstration projects which have been supported in the framework of the energy demonstration programme since 1983. It includes reports by each of the contractors who submitted proposals in 1987 and 1988, describing progress within their projects. Projects accepted from earlier calls for proposals and not yet completed were reviewed by a rapporteur and are discussed in the summary section. The results of the performance monitoring of all projects and the lessons drawn from the practical experience of the projects are also presented in the summaries and conclusions. Contractors whose projects were submitted in 1989 were also present at the meeting and contributed to the reported discussions. This proceeding is divided into four sessions (General, Housing, technical presentations, other applications) and 24 papers are offered.

  8. AVNG system demonstration

    Thron, Jonathan Louis [Los Alamos National Laboratory; Mac Arthur, Duncan W [Los Alamos National Laboratory; Kondratov, Sergey [VNIIEF; Livke, Alexander [VNIIEF; Razinkov, Sergey [VNIIEF

    2010-01-01

    An attribute measurement system (AMS) measures a number of unclassified attributes of potentially classified material. By only displaying these unclassified results as red or green lights, the AMS protects potentially classified information while still generating confidence in the measurement result. The AVNG implementation that we describe is an AMS built by RFNC - VNIIEF in Sarov, Russia. To provide additional confidence, the AVNG was designed with two modes of operation. In the secure mode, potentially classified measurements can be made with only the simple red light/green light display. In the open mode, known unclassified material can be measured with complete display of the information collected from the radiation detectors. The AVNG demonstration, which occurred in Sarov, Russia in June 2009 for a joint US/Russian audience, included exercising both modes of AVNG operation using a number of multi-kg plutonium sources. In addition to describing the demonstration, we will show photographs and/or video taken of AVNG operation.

  9. Software system safety

    Uber, James G.

    1988-01-01

    Software itself is not hazardous, but since software and hardware share common interfaces there is an opportunity for software to create hazards. Further, these software systems are complex, and proven methods for the design, analysis, and measurement of software safety are not yet available. Some past software failures, future NASA software trends, software engineering methods, and tools and techniques for various software safety analyses are reviewed. Recommendations to NASA are made based on this review.

  10. Antares: preliminary demonstrator results

    Kouchner, A.

    2000-05-01

    The ANTARES collaboration is building an undersea neutrino telescope off Toulon (Mediterranean sea) with effective area ∼ 0.1 km 2 . An extensive study of the site properties has been achieved together with software analysis in order to optimize the performance of the detector. Results are summarized here. An instrumented line, linked to shore for first time via an electro-optical cable, has been immersed late 1999. The preliminary results of this demonstrator line are reported. (author)

  11. The Majorana Demonstrator

    Aguayo, Estanislao; Fast, James E.; Hoppe, Eric W.; Keillor, Martin E.; Kephart, Jeremy D.; Kouzes, Richard T.; LaFerriere, Brian D.; Merriman, Jason H.; Orrell, John L.; Overman, Nicole R.; Avignone, Frank T.; Back, Henning O.; Combs, Dustin C.; Leviner, L.; Young, A.; Barabash, Alexander S.; Konovalov, S.; Vanyushin, I.; Yumatov, Vladimir; Bergevin, M.; Chan, Yuen-Dat; Detwiler, Jason A.; Loach, J. C.; Martin, R. D.; Poon, Alan; Prior, Gersende; Vetter, Kai; Bertrand, F.; Cooper, R. J.; Radford, D. C.; Varner, R. L.; Yu, Chang-Hong; Boswell, M.; Elliott, S.; Gehman, Victor M.; Hime, Andrew; Kidd, M. F.; LaRoque, B. H.; Rielage, Keith; Ronquest, M. C.; Steele, David; Brudanin, V.; Egorov, Viatcheslav; Gusey, K.; Kochetov, Oleg; Shirchenko, M.; Timkin, V.; Yakushev, E.; Busch, Matthew; Esterline, James H.; Tornow, Werner; Christofferson, Cabot-Ann; Horton, Mark; Howard, S.; Sobolev, V.; Collar, J. I.; Fields, N.; Creswick, R.; Doe, Peter J.; Johnson, R. A.; Knecht, A.; Leon, Jonathan D.; Marino, Michael G.; Miller, M. L.; Robertson, R. G. H.; Schubert, Alexis G.; Wolfe, B. A.; Efremenko, Yuri; Ejiri, H.; Hazama, R.; Nomachi, Masaharu; Shima, T.; Finnerty, P.; Fraenkle, Florian; Giovanetti, G. K.; Green, M.; Henning, Reyco; Howe, M. A.; MacMullin, S.; Phillips, D.; Snavely, Kyle J.; Strain, J.; Vorren, Kris R.; Guiseppe, Vincente; Keller, C.; Mei, Dong-Ming; Perumpilly, Gopakumar; Thomas, K.; Zhang, C.; Hallin, A. L.; Keeter, K.; Mizouni, Leila; Wilkerson, J. F.

    2011-09-03

    A brief review of the history and neutrino physics of double beta decay is given. A description of the MAJORANA DEMONSTRATOR research and development program, including background reduction techniques, is presented in some detail. The application of point contact (PC) detectors to the experiment is discussed, including the effectiveness of pulse shape analysis. The predicted sensitivity of a PC detector array enriched to 86% to 76Ge is given.

  12. IGCC technology and demonstration

    Palonen, J [A. Ahlstrom Corporation, Karhula (Finland). Hans Ahlstrom Lab.; Lundqvist, R G [A. Ahlstrom Corporation, Helsinki (Finland); Staahl, K [Sydkraft AB, Malmoe (Sweden)

    1997-12-31

    Future energy production will be performed by advanced technologies that are more efficient, more environmentally friendly and less expensive than current technologies. Integrated gasification combined cycle (IGCC) power plants have been proposed as one of these systems. Utilising biofuels in future energy production will also be emphasised since this lowers substantially carbon dioxide emissions into the atmosphere due to the fact that biomass is a renewable form of energy. Combining advanced technology and biomass utilisation is for this reason something that should and will be encouraged. A. Ahlstrom Corporation of Finland and Sydkraft AB of Sweden have as one part of company strategies adopted this approach for the future. The companies have joined their resources in developing a biomass-based IGCC system with the gasification part based on pressurised circulating fluidized-bed technology. With this kind of technology electrical efficiency can be substantially increased compared to conventional power plants. As a first concrete step, a decision has been made to build a demonstration plant. This plant, located in Vaernamo, Sweden, has already been built and is now in commissioning and demonstration stage. The system comprises a fuel drying plant, a pressurised CFB gasifier with gas cooling and cleaning, a gas turbine, a waste heat recovery unit and a steam turbine. The plant is the first in the world where the integration of a pressurised gasifier with a gas turbine will be realised utilising a low calorific gas produced from biomass. The capacity of the Vaernamo plant is 6 MW of electricity and 9 MW of district heating. Technology development is in progress for design of plants of sizes from 20 to 120 MWe. The paper describes the Bioflow IGCC system, the Vaernamo demonstration plant and experiences from the commissioning and demonstration stages. (orig.)

  13. IGCC technology and demonstration

    Palonen, J. [A. Ahlstrom Corporation, Karhula (Finland). Hans Ahlstrom Lab.; Lundqvist, R.G. [A. Ahlstrom Corporation, Helsinki (Finland); Staahl, K. [Sydkraft AB, Malmoe (Sweden)

    1996-12-31

    Future energy production will be performed by advanced technologies that are more efficient, more environmentally friendly and less expensive than current technologies. Integrated gasification combined cycle (IGCC) power plants have been proposed as one of these systems. Utilising biofuels in future energy production will also be emphasised since this lowers substantially carbon dioxide emissions into the atmosphere due to the fact that biomass is a renewable form of energy. Combining advanced technology and biomass utilisation is for this reason something that should and will be encouraged. A. Ahlstrom Corporation of Finland and Sydkraft AB of Sweden have as one part of company strategies adopted this approach for the future. The companies have joined their resources in developing a biomass-based IGCC system with the gasification part based on pressurised circulating fluidized-bed technology. With this kind of technology electrical efficiency can be substantially increased compared to conventional power plants. As a first concrete step, a decision has been made to build a demonstration plant. This plant, located in Vaernamo, Sweden, has already been built and is now in commissioning and demonstration stage. The system comprises a fuel drying plant, a pressurised CFB gasifier with gas cooling and cleaning, a gas turbine, a waste heat recovery unit and a steam turbine. The plant is the first in the world where the integration of a pressurised gasifier with a gas turbine will be realised utilising a low calorific gas produced from biomass. The capacity of the Vaernamo plant is 6 MW of electricity and 9 MW of district heating. Technology development is in progress for design of plants of sizes from 20 to 120 MWe. The paper describes the Bioflow IGCC system, the Vaernamo demonstration plant and experiences from the commissioning and demonstration stages. (orig.)

  14. Lunar Water Resource Demonstration

    Muscatello, Anthony C.

    2008-01-01

    In cooperation with the Canadian Space Agency, the Northern Centre for Advanced Technology, Inc., the Carnegie-Mellon University, JPL, and NEPTEC, NASA has undertaken the In-Situ Resource Utilization (ISRU) project called RESOLVE. This project is a ground demonstration of a system that would be sent to explore permanently shadowed polar lunar craters, drill into the regolith, determine what volatiles are present, and quantify them in addition to recovering oxygen by hydrogen reduction. The Lunar Prospector has determined these craters contain enhanced hydrogen concentrations averaging about 0.1%. If the hydrogen is in the form of water, the water concentration would be around 1%, which would translate into billions of tons of water on the Moon, a tremendous resource. The Lunar Water Resource Demonstration (LWRD) is a part of RESOLVE designed to capture lunar water and hydrogen and quantify them as a backup to gas chromatography analysis. This presentation will briefly review the design of LWRD and some of the results of testing the subsystem. RESOLVE is to be integrated with the Scarab rover from CMIJ and the whole system demonstrated on Mauna Kea on Hawaii in November 2008. The implications of lunar water for Mars exploration are two-fold: 1) RESOLVE and LWRD could be used in a similar fashion on Mars to locate and quantify water resources, and 2) electrolysis of lunar water could provide large amounts of liquid oxygen in LEO, leading to lower costs for travel to Mars, in addition to being very useful at lunar outposts.

  15. Waste and Disposal: Demonstration

    Neerdael, B.; Buyens, M.; De Bruyn, D.; Volckaert, G.

    2002-01-01

    Within the Belgian R and D programme on geological disposal, demonstration experiments have become increasingly important. In this contribution to the scientific report 2001, an overview is given of SCK-CEN's activities and achievements in the field of large-scale demonstration experiments. In 2001, main emphasis was on the PRACLAY project, which is a large-scale experiment to demonstrate the construction and the operation of a gallery for the disposal of HLW in a clay formation. The PRACLAY experiment will contribute to enhance understanding of water flow and mass transport in dense clay-based materials as well as to improve the design of the reference disposal concept. In the context of PRACLAY, a surface experiment (OPHELIE) has been developed to prepare and to complement PRACLAY-related experimental work in the HADES Underground Research Laboratory. In 2001, efforts were focussed on the operation of the OPHELIE mock-up. SCK-CEN also contributed to the SELFRAC roject which studies the self-healing of fractures in a clay formation

  16. Demonstration and Dialogue: Mediation in Swedish Nuclear Waste Management. Deliverable D10

    Elam, Mark; Sundqvist, Goeran; Lidberg, Maria; Soneryd, Linda

    2008-10-01

    This report analyses mediation and mediators in Swedish nuclear waste management. Mediation is about establishing agreement and building common knowledge. It is argued that demonstrations and dialogue are the two prominent approaches to mediation in Swedish nuclear waste management. Mediation through demonstration is about showing, displaying, and pointing out a path to safe disposal for inspection. It implies a strict division between demonstrator and audience. Mediation through dialogue on the other hand, is about collective acknowledgements of uncertainty and suspensions of judgement creating room for broader discussion. In Sweden, it is the Swedish Nuclear Fuel and Waste Management Co. (SKB) that is tasked with finding a method and a site for the final disposal of the nation's nuclear waste. Two different legislative frameworks cover this process. In accordance with the Act on Nuclear Activities, SKB is required to demonstrate the safety of its planned nuclear waste management system to the government, while in respect of the Swedish Environmental Code, they are obliged to organize consultations with the public. How SKB combines these requirements is the main question under investigation in this report in relation to materials deriving from three empirical settings: 1) SKB's safety analyses, 2) SKB's public consultation activities and 3) the 'dialogue projects', initiated by other actors than SKB broadening the public arena for discussion. In conclusion, an attempt is made to characterise the long-term interplay of demonstration and dialogue in Swedish nuclear waste management

  17. Reactor safety

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  18. China's nuclear safety regulatory body: The national nuclear safety administration

    Zhang Shiguan

    1991-04-01

    The establishment of an independent nuclear safety regulatory body is necessary for ensuring the safety of nuclear installations and nuclear fuel. Therefore the National Nuclear Safety Administration was established by the state. The aim, purpose, organization structure and main tasks of the Administration are presented. At the same time the practical examples, such as nuclear safety regulation on the Qinshan Nuclear Power Plant, safety review and inspections for the Daya Bay Nuclear Power Plant during the construction, and nuclear material accounting and management system in the nuclear fuel fabrication plant in China, are given in order to demonstrate the important roles having been played on nuclear safety by the Administration after its founding

  19. K Basin safety analysis

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

    1994-01-01

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

  20. Towards confidence in transport safety

    Robison, R.W.

    1992-01-01

    The U.S. Department of Energy (US DOE) plans to demonstrate to the public that high-level waste can be transported safely to the proposed repository. The author argues US DOE should begin now to demonstrate its commitment to safety by developing an extraordinary safety program for nuclear cargo it is now shipping. The program for current shipments should be developed with State, Tribal, and local officials. Social scientists should be involved in evaluating the effect of the safety program on public confidence. The safety program developed in cooperation with western states for shipments to the Waste Isolation Pilot plant is a good basis for designing that extraordinary safety program

  1. Aerospace Communications Security Technologies Demonstrated

    Griner, James H.; Martzaklis, Konstantinos S.

    2003-01-01

    In light of the events of September 11, 2001, NASA senior management requested an investigation of technologies and concepts to enhance aviation security. The investigation was to focus on near-term technologies that could be demonstrated within 90 days and implemented in less than 2 years. In response to this request, an internal NASA Glenn Research Center Communications, Navigation, and Surveillance Aviation Security Tiger Team was assembled. The 2-year plan developed by the team included an investigation of multiple aviation security concepts, multiple aircraft platforms, and extensively leveraged datalink communications technologies. It incorporated industry partners from NASA's Graphical Weather-in-the-Cockpit research, which is within NASA's Aviation Safety Program. Two concepts from the plan were selected for demonstration: remote "black box," and cockpit/cabin surveillance. The remote "black box" concept involves real-time downlinking of aircraft parameters for remote monitoring and archiving of aircraft data, which would assure access to the data following the loss or inaccessibility of an aircraft. The cockpit/cabin surveillance concept involves remote audio and/or visual surveillance of cockpit and cabin activity, which would allow immediate response to any security breach and would serve as a possible deterrent to such breaches. The datalink selected for the demonstrations was VDL Mode 2 (VHF digital link), the first digital datalink for air-ground communications designed for aircraft use. VDL Mode 2 is beginning to be implemented through the deployment of ground stations and aircraft avionics installations, with the goal of being operational in 2 years. The first demonstration was performed December 3, 2001, onboard the LearJet 25 at Glenn. NASA worked with Honeywell, Inc., for the broadcast VDL Mode 2 datalink capability and with actual Boeing 757 aircraft data. This demonstration used a cockpitmounted camera for video surveillance and a coupling to

  2. Demonstration of HITEX

    Morrison, H.D.; Woodall, K.B.

    1993-01-01

    A model reactor for HITEX successfully demonstrated the concept of high-temperature isotopic exchange in a closed loop simulating the conditions for fusion fuel cleanup. The catalyst of platinum on alumina pellets provided a surface area large enough to operate the reactor at 400 degrees celsius with flow rates up to 2 L/min. A 15-L tank containing a mixture of 4% CD 4 in H 2 was depleted in deuterium within 75 minutes down to 100 ppm HD above the natural concentration of HD in the make-up hydrogen stream. The application to tritium removal from tritiated impurities in a hydrogen stream will work as well or better

  3. Visual Electricity Demonstrator

    Lincoln, James

    2017-09-01

    The Visual Electricity Demonstrator (VED) is a linear diode array that serves as a dynamic alternative to an ammeter. A string of 48 red light-emitting diodes (LEDs) blink one after another to create the illusion of a moving current. Having the current represented visually builds an intuitive and qualitative understanding about what is happening in a circuit. In this article, I describe several activities for this device and explain how using this technology in the classroom can enhance the understanding and appreciation of physics.

  4. Exploration Medical System Demonstration

    Rubin, D. A.; Watkins, S. D.

    2014-01-01

    BACKGROUND: Exploration class missions will present significant new challenges and hazards to the health of the astronauts. Regardless of the intended destination, beyond low Earth orbit a greater degree of crew autonomy will be required to diagnose medical conditions, develop treatment plans, and implement procedures due to limited communications with ground-based personnel. SCOPE: The Exploration Medical System Demonstration (EMSD) project will act as a test bed on the International Space Station (ISS) to demonstrate to crew and ground personnel that an end-to-end medical system can assist clinician and non-clinician crew members in optimizing medical care delivery and data management during an exploration mission. Challenges facing exploration mission medical care include limited resources, inability to evacuate to Earth during many mission phases, and potential rendering of medical care by non-clinicians. This system demonstrates the integration of medical devices and informatics tools for managing evidence and decision making and can be designed to assist crewmembers in nominal, non-emergent situations and in emergent situations when they may be suffering from performance decrements due to environmental, physiological or other factors. PROJECT OBJECTIVES: The objectives of the EMSD project are to: a. Reduce or eliminate the time required of an on-orbit crew and ground personnel to access, transfer, and manipulate medical data. b. Demonstrate that the on-orbit crew has the ability to access medical data/information via an intuitive and crew-friendly solution to aid in the treatment of a medical condition. c. Develop a common data management framework that can be ubiquitously used to automate repetitive data collection, management, and communications tasks for all activities pertaining to crew health and life sciences. d. Ensure crew access to medical data during periods of restricted ground communication. e. Develop a common data management framework that

  5. Commercial incineration demonstration

    Vavruska, J.S.; Borduin, L.C.

    1982-01-01

    Low-level radioactive wastes (LLW) generated by nuclear utilities presently are shipped to commercial burial grounds for disposal. Increasing transportation and disposal costs have caused industry to consider incineration as a cost-effective means of volume reduction of combustible LLW. Repeated inquiries from the nuclear industry regarding the applicability of the Los Alamos controlled air incineration (CAI) design led the DOE to initiate a commercial demonstration program in FY-1980. Development studies and results in support of this program involving ion exchange resin incineration and fission/activation product distributions within the Los Alamos CAI are described

  6. Demonstration tokamak power plant

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  7. Risk analyses of nuclear power plants

    Jehee, J.N.T.; Seebregts, A.J.

    1991-02-01

    Probabilistic risk analyses of nuclear power plants are carried out by systematically analyzing the possible consequences of a broad spectrum of causes of accidents. The risk can be expressed in the probabilities for melt down, radioactive releases, or harmful effects for the environment. Following risk policies for chemical installations as expressed in the mandatory nature of External Safety Reports (EVRs) or, e.g., the publication ''How to deal with risks'', probabilistic risk analyses are required for nuclear power plants

  8. Measurement equivalence of patient safety climate in Chinese hospitals: can we compare across physicians and nurses?

    Zhu, Junya

    2018-06-11

    Self-report instruments have been widely used to better understand variations in patient safety climate between physicians and nurses. Research is needed to determine whether differences in patient safety climate reflect true differences in the underlying concepts. This is known as measurement equivalence, which is a prerequisite for meaningful group comparisons. This study aims to examine the degree of measurement equivalence of the responses to a patient safety climate survey of Chinese hospitals and to demonstrate how the measurement equivalence method can be applied to self-report climate surveys for patient safety research. Using data from the Chinese Hospital Survey of Patient Safety Climate from six Chinese hospitals in 2011, we constructed two groups: physicians and nurses (346 per group). We used multiple-group confirmatory factor analyses to examine progressively more stringent restrictions for measurement equivalence. We identified weak factorial equivalence across the two groups. Strong factorial equivalence was found for Organizational Learning, Unit Management Support for Safety, Adequacy of Safety Arrangements, Institutional Commitment to Safety, Error Reporting and Teamwork. Strong factorial equivalence, however, was not found for Safety System, Communication and Peer Support and Staffing. Nevertheless, further analyses suggested that nonequivalence did not meaningfully affect the conclusions regarding physician-nurse differences in patient safety climate. Our results provide evidence of at least partial equivalence of the survey responses between nurses and physicians, supporting mean comparisons of its constructs between the two groups. The measurement equivalence approach is essential to ensure that conclusions about group differences are valid.

  9. Smart Grid Demonstration Project

    Miller, Craig [National Rural Electric Cooperative Association, Arlington, VA (United States); Carroll, Paul [National Rural Electric Cooperative Association, Arlington, VA (United States); Bell, Abigail [National Rural Electric Cooperative Association, Arlington, VA (United States)

    2015-03-11

    The National Rural Electric Cooperative Association (NRECA) organized the NRECA-U.S. Department of Energy (DOE) Smart Grid Demonstration Project (DE-OE0000222) to install and study a broad range of advanced smart grid technologies in a demonstration that spanned 23 electric cooperatives in 12 states. More than 205,444 pieces of electronic equipment and more than 100,000 minor items (bracket, labels, mounting hardware, fiber optic cable, etc.) were installed to upgrade and enhance the efficiency, reliability, and resiliency of the power networks at the participating co-ops. The objective of this project was to build a path for other electric utilities, and particularly electrical cooperatives, to adopt emerging smart grid technology when it can improve utility operations, thus advancing the co-ops’ familiarity and comfort with such technology. Specifically, the project executed multiple subprojects employing a range of emerging smart grid technologies to test their cost-effectiveness and, where the technology demonstrated value, provided case studies that will enable other electric utilities—particularly electric cooperatives— to use these technologies. NRECA structured the project according to the following three areas: Demonstration of smart grid technology; Advancement of standards to enable the interoperability of components; and Improvement of grid cyber security. We termed these three areas Technology Deployment Study, Interoperability, and Cyber Security. Although the deployment of technology and studying the demonstration projects at coops accounted for the largest portion of the project budget by far, we see our accomplishments in each of the areas as critical to advancing the smart grid. All project deliverables have been published. Technology Deployment Study: The deliverable was a set of 11 single-topic technical reports in areas related to the listed technologies. Each of these reports has already been submitted to DOE, distributed to co-ops, and

  10. Safety climate and attitude as evaluation measures of organizational safety.

    Isla Díaz, R; Díaz Cabrera, D

    1997-09-01

    The main aim of this research is to develop a set of evaluation measures for safety attitudes and safety climate. Specifically it is intended: (a) to test the instruments; (b) to identify the essential dimensions of the safety climate in the airport ground handling companies; (c) to assess the quality of the differences in the safety climate for each company and its relation to the accident rate; (d) to analyse the relationship between attitudes and safety climate; and (e) to evaluate the influences of situational and personal factors on both safety climate and attitude. The study sample consisted of 166 subjects from three airport companies. Specifically, this research was centered on ground handling departments. The factor analysis of the safety climate instrument resulted in six factors which explained 69.8% of the total variance. We found significant differences in safety attitudes and climate in relation to type of enterprise.

  11. Safety Culture Survey in Krsko NPP

    Strucic, M.; Bilic Zadric, T.

    2008-01-01

    The high level of nuclear safety, stability and competitiveness of electricity production, and public acceptability are the main objectives of Krsko Nuclear Power Plant. This is achievable only in environment where strong Safety Culture is taking dominant place in the way how employees communicate, perform tasks, share their ideas and attitudes, and demonstrate their concern in all aspects of work and coexistence. To achieve these objectives, behaviour of all employees as well as specific ethical values must become more transparent and that must arise from the heart of organization. Continuous ongoing and periodic self assessments of Safety Culture in Krsko NPP present major tools in implementation process of this approach. Benefits from Periodic interdisciplinary focused self assessment approach, which main intention is finding the strengths and potential areas for improvements, was used second time to assess the area of Safety Culture in Krsko NPP. Main objectives of self assessment, performed in 2006, were 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. For the purpose of effective self assessment, extensive questionnaire was used to obtain information that is representative for whole organization. Wide range of questions was chosen to cover five major characteristics of safety culture: Accountability for safety is clear, Safety is integrated into all activities, Safety culture is learning-driven, Leadership for safety is clear and Safety is a clearly recognized value. 484 Krsko NPP employees and 96 contractors were participated in survey. 70-question survey provided information that was quantified and results compared between groups. Anonymity of participant, as well as their willingness to contribute in this assessment implicates the high level of their openness in answering the questions. High number of participant made analysis of

  12. Electrodynamic Dust Shield Demonstrator

    Stankie, Charles G.

    2013-01-01

    The objective of the project was to design and manufacture a device to demonstrate a new technology developed by NASA's Electrostatics and Surface Physics Laboratory. The technology itself is a system which uses magnetic principles to remove regolith dust from its surface. This project was to create an enclosure that will be used to demonstrate the effectiveness of the invention to The Office of the Chief Technologist. ONE of the most important challenges of space exploration is actually caused by something very small and seemingly insignificant. Dust in space, most notably on the moon and Mars, has caused many unforeseen issues. Dirt and dust on Earth, while a nuisance, can be easily cleaned and kept at bay. However, there is considerably less weathering and erosion in space. As a result, the microscopic particles are extremely rough and abrasive. They are also electrostatically charged, so they cling to everything they make contact with. This was first noted to be a major problem during the Apollo missions. Dust would stick to the spacesuits, and could not be wiped off as predicted. Dust was brought back into the spacecraft, and was even inhaled by astronauts. This is a major health hazard. Atmospheric storms and other events can also cause dust to coat surfaces of spacecraft. This can cause abrasive damage to the craft. The coating can also reduce the effectiveness of thermal insulation and solar panels.' A group of engineers at Kennedy Space Center's Electrostatics and Surface Physics Laboratory have developed a new technology, called the Electrodynamic Dust Shield, to help alleviate these problems. It is based off of the electric curtain concept developed at NASA in 1967. "The EDS is an active dust mitigation technology that uses traveling electric fields to transport electrostatically charged dust particles along surfaces. To generate the traveling electric fields, the EDS consists of a multilayer dielectric coating with an embedded thin electrode grid

  13. Fuel Cell Demonstration Program

    Gerald Brun

    2006-09-15

    In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance

  14. Safety assessment as basis for the decision making process

    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

  15. TRAC analyses for CCTF and SCTF tests and UPTF design/operation

    Williams, K.A.

    1983-01-01

    The 2D/3D Program is a multinational (Germany, Japan, and the United States) experimental and analytical nuclear reactor safety research program. The Los Alamos analysis effort is functioning as a vital part of the 2D/3D program. The CCTF and SCTF analyses have demonstrated that TRAC-PF1 can correctly predict multidimensional, nonequilibrium behavior in large-scale facilities prototypical of actual PWR's. Through these and future TRAC analyses the experimental findings can be related from facility to facility, and the results of this research program can be directly related to licensing concerns affecting actual PWR's

  16. Fusion-power demonstration

    Henning, C.D.; Logan, B.G.; Carlson, G.A.; Neef, W.S.; Moir, R.W.; Campbell, R.B.; Botwin, R.; Clarkson, I.R.; Carpenter, T.J.

    1983-01-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

  17. Spent fuel pyroprocessing demonstration

    McFarlane, L.F.; Lineberry, M.J.

    1995-01-01

    A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

  18. Industrial demonstration trials

    Gelee, M.; Fabre, C.; Villepoix, R. de; Fra, J.; Le Foulgoc, L.; Morel, Y.; Querite, P.; Roques, R.

    1975-01-01

    Prototypes of the plant components, meeting the specifications set by the process and built by industrial firms in collaboration with the supervisor and the C.E.A., are subjected to trial runs on the UF 6 test bench of the Pierrelatte testing zone. These items of equipment (diffuser, compressor, exchanger) are placed in an industrial operation context very similar to that of an enrichment plant. Their performance is measured within a broad region around the working point and their reliability observed over periods up to several tens of thousands of hours. Between 1969 and 1973 six industrial demonstration test benches have been built, marking the stages in the technical preparation of the 1973 file on the basis of which the decision of building was taken by Eurodif [fr

  19. Fusion Power Demonstration III

    Lee, J.D.

    1985-07-01

    This is the third in the series of reports covering the Fusion Power Demonstration (FPD) design study. This volume considers the FPD-III configuration that incorporates an octopole end plug. As compared with the quadrupole end-plugged designs of FPD-I and FPD-II, this octopole configuration reduces the number of end cell magnets and shortens the minimum ignition length of the central cell. The end-cell plasma length is also reduced, which in turn reduces the size and cost of the end cell magnets and shielding. As a contiuation in the series of documents covering the FPD, this report does not stand alone as a design description of FPD-III. Design details of FPD-III subsystems that do not differ significantly from those of the FPD-II configuration are not duplicated in this report

  20. TPA device for demonstration

    1980-02-01

    The TPA (torus plasma for amature) is a small race-trac type device made by the technical service division to demonstrate basic properties of plasma such as electron temperature, conductivity, effect of helical field for toroidal drift, and shape of plasma in mirror and cusp magnetic field in linear section. The plasmas are produced by RF discharge (-500W) and/or DC discharge (-30 mA) within glass discharge tube. Where major radius is 50 cm, length of linear section is 50 cm, toroidal magnetic field is 200 gauss. The device has been designed to be compact with only 100 V power source (-3.2 KW for the case without helical field) and to be full automatic sequence of operation. (author)

  1. Fusion power demonstration

    Henning, C.D.; Logan, B.G.

    1983-01-01

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

  2. Dynamic wall demonstration project

    Nakatsui, L.; Mayhew, W.

    1990-12-01

    The dynamic wall concept is a ventilation strategy that can be applied to a single family dwelling. With suitable construction, outside air can be admitted through the exterior walls of the house to the interior space to function as ventilation air. The construction and performance monitoring of a demonstration house built to test the dynamic wall concept in Sherwood Park, Alberta, is described. The project had the objectives of demonstrating and assessing the construction methods; determining the cost-effectiveness of the concept in Alberta; analyzing the operation of the dynamic wall system; and determining how other components and systems in the house interact with the dynamic wall. The exterior wall construction consisted of vinyl siding, spun-bonded polyolefin-backed (SBPO) rigid fiberglass sheathing, 38 mm by 89 mm framing, fiberglass batt insulation and 12.7 mm drywall. The mechanical system was designed to operate in the dynamic (negative pressure) mode, however flexibility was provided to allow operation in the static (balanced pressure) mode to permit monitoring of the walls as if they were in a conventional house. The house was monitored by an extensive computerized monitoring system. Dynamic wall operation was dependent on pressure and temperature differentials between indoor and outdoor as well as wind speed and direction. The degree of heat gain was found to be ca 74% of the indoor-outdoor temperature differential. Temperature of incoming dynamic air was significantly affected by solar radiation and measurement of indoor air pollutants found no significant levels. 4 refs., 34 figs., 11 tabs.

  3. PWR reload safety evaluation methodology

    Doshi, P.K.; Chapin, D.L.; Love, D.S.

    1993-01-01

    The current practice for WWER safety analysis is to prepare the plant Safety Analysis Report (SAR) for initial plant operation. However, the existing safety analysis is typically not evaluated for reload cycles to confirm that all safety limits are met. In addition, there is no systematic reanalysis or reevaluation of the safety analyses after there have been changes made to the plant. The Westinghouse process is discussed which is in contrast to this and in which the SAR conclusions are re-validated through evaluation and/or analysis of each reload cycle. (Z.S.)

  4. Developing tools for the safety specification in risk management plans: lessons learned from a pilot project.

    Cooper, Andrew J P; Lettis, Sally; Chapman, Charlotte L; Evans, Stephen J W; Waller, Patrick C; Shakir, Saad; Payvandi, Nassrin; Murray, Alison B

    2008-05-01

    Following the adoption of the ICH E2E guideline, risk management plans (RMP) defining the cumulative safety experience and identifying limitations in safety information are now required for marketing authorisation applications (MAA). A collaborative research project was conducted to gain experience with tools for presenting and evaluating data in the safety specification. This paper presents those tools found to be useful and the lessons learned from their use. Archive data from a successful MAA were utilised. Methods were assessed for demonstrating the extent of clinical safety experience, evaluating the sensitivity of the clinical trial data to detect treatment differences and identifying safety signals from adverse event and laboratory data to define the extent of safety knowledge with the drug. The extent of clinical safety experience was demonstrated by plots of patient exposure over time. Adverse event data were presented using dot plots, which display the percentages of patients with the events of interest, the odds ratio, and 95% confidence interval. Power and confidence interval plots were utilised for evaluating the sensitivity of the clinical database to detect treatment differences. Box and whisker plots were used to display laboratory data. This project enabled us to identify new evidence-based methods for presenting and evaluating clinical safety data. These methods represent an advance in the way safety data from clinical trials can be analysed and presented. This project emphasises the importance of early and comprehensive planning of the safety package, including evaluation of the use of epidemiology data.

  5. Functional Safety Specification of Communication Profile PROFIsafe

    Jan Rofar

    2006-01-01

    Full Text Available Paper maps the trends in area of safety-related communication within PROFIBUS and PROFINET industry networks. There are analyses safety measures and Fail-safe parameters of PROFIsafe profile in version V2 and their localisation in Safety Communication Layer SCL, which guarantees Safety Integrity Level SIL according to standard IEC 61508. The last chapter analyses the reaction in the event of fault during transmission of messages.

  6. The Effect of Age, Parity and Body Mass Index on the Efficacy, Safety, Placement and User Satisfaction Associated With Two Low-Dose Levonorgestrel Intrauterine Contraceptive Systems: Subgroup Analyses of Data From a Phase III Trial.

    Kristina Gemzell-Danielsson

    Full Text Available Two low-dose levonorgestrel intrauterine contraceptive systems (LNG-IUSs; total content 13.5 mg [average approx. 8 μg/24 hours over the first year; LNG-IUS 8] and total content 19.5 mg [average approx. 13 μg/24 hours over the first year; LNG-IUS 13] have previously been shown to be highly effective (3-year Pearl Indices: 0.33 and 0.31, respectively, safe and well tolerated. The present subgroup analyses evaluated whether or not outcomes were affected by parity, age (18-25 vs 26-35 years, or body mass index (BMI, <30 vs ≥30 kg/m2.Nulliparous and parous women aged 18‒35 years with regular menstrual cycles (21‒35 days requesting contraception were randomized to 3 years of LNG-IUS 8 or LNG-IUS 13 use.In the LNG-IUS 8 and LNG-IUS 13 groups, 1432 and 1452 women, respectively, had a placement attempted and were included in the full analysis set; 39.2%, 39.2% and 17.1% were 18-25 years old, nulliparous and had a BMI ≥30 kg/m2, respectively. Both systems were similarly effective regardless of age, parity or BMI; the subgroup Pearl Indices had widely overlapping 95% confidence intervals. Placement of LNG-IUS 8 and LNG-IUS 13 was easier (p < 0.0001 and less painful (p < 0.0001 in women who had delivered vaginally than in women who had not. The complete/partial expulsion rate was 2.2-4.2% across all age and parity subgroups and higher in parous than in nulliparous women (p = 0.004. The incidence of pelvic inflammatory disease was 0.1-0.6% across all age and parity subgroups: nulliparous and younger women were not at higher risk than parous and older women, respectively. The ectopic pregnancy rate was 0.3-0.4% across all age and parity subgroups. Across all age and parity subgroups, the 3-year completion rate was 50.9-61.3% for LNG-IUS 8 and 57.9-61.1% for LNG-IUS 13, and was higher (p = 0.0001 among older than younger women in the LNG-IUS 8 group only.LNG-IUS 8 and LNG-IUS 13 were highly effective, safe and well tolerated regardless of age or

  7. Safety analysis for 'Fugen'

    1997-10-01

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

  8. A Demonstration of Lusail

    Mansour, Essam; Abdelaziz, Ibrahim; Ouzzani, Mourad; Aboulnaga, Ashraf; Kalnis, Panos

    2017-01-01

    There has been a proliferation of datasets available as interlinked RDF data accessible through SPARQL endpoints. This has led to the emergence of various applications in life science, distributed social networks, and Internet of Things that need to integrate data from multiple endpoints. We will demonstrate Lusail; a system that supports the need of emerging applications to access tens to hundreds of geo-distributed datasets. Lusail is a geo-distributed graph engine for querying linked RDF data. Lusail delivers outstanding performance using (i) a novel locality-aware query decomposition technique that minimizes the intermediate data to be accessed by the subqueries, and (ii) selectivityawareness and parallel query execution to reduce network latency and to increase parallelism. During the demo, the audience will be able to query actually deployed RDF endpoints as well as large synthetic and real benchmarks that we have deployed in the public cloud. The demo will also show that Lusail outperforms state-of-the-art systems by orders of magnitude in terms of scalability and response time.

  9. A Demonstration of Lusail

    Mansour, Essam

    2017-05-10

    There has been a proliferation of datasets available as interlinked RDF data accessible through SPARQL endpoints. This has led to the emergence of various applications in life science, distributed social networks, and Internet of Things that need to integrate data from multiple endpoints. We will demonstrate Lusail; a system that supports the need of emerging applications to access tens to hundreds of geo-distributed datasets. Lusail is a geo-distributed graph engine for querying linked RDF data. Lusail delivers outstanding performance using (i) a novel locality-aware query decomposition technique that minimizes the intermediate data to be accessed by the subqueries, and (ii) selectivityawareness and parallel query execution to reduce network latency and to increase parallelism. During the demo, the audience will be able to query actually deployed RDF endpoints as well as large synthetic and real benchmarks that we have deployed in the public cloud. The demo will also show that Lusail outperforms state-of-the-art systems by orders of magnitude in terms of scalability and response time.

  10. Demonstration exercise 'Cavtat 09'

    Trut, D.

    2009-01-01

    The demonstration exercise is to show a terrorist attack in urban area resulting in a certain number of injured people. On 7th April 2009 a terrorist group HAL 9000 is in Cavtat and set up an explosive devices with chemical reagents in several spots with intention to activate them and cause great number of victims. On the same day, in area of the Cavtat Croatia Hotel, which is hosting the world CBMTS Congress, Cavtat Police Station notice several masked persons, in escapement. Hotel personnel alerted the County 112 Center about noticed devices placed by chlorine dioxide tanks, for water conditioning. Intervention police came to block entrance to this area and evacuate hotel's guests and congress members. An explosion and fire occurs from where the position of water-conditioning plant and chlorine dioxide tank. The 112 Center alarms fire-fighters for fight fire and decontamination action and HAZMAT Civil Support Team from Georgia (participated the congress). In the meantime, guests have been instructed not to leave their rooms and to hermetically close doors and windows with available material to keep away potential toxic fume. Decision makers form the County Protection and Rescue Headquarters monitors the situation till the end of alert for the population in the area of Cavtat.(author)

  11. Tidd PFBC demonstration project

    Marrocco, M. [American Electric Power, Columbus, OH (United States)

    1997-12-31

    The Tidd project was one of the first joint government-industry ventures to be approved by the US Department of Energy (DOE) in its Clean Coal Technology Program. In March 1987, DOE signed an agreement with the Ohio Power Company, a subsidiary of American Electric Power, to refurbish the then-idle Tidd plant on the banks of the Ohio River with advanced pressurized fluidized bed technology. Testing ended after 49 months of operation, 100 individual tests, and the generation of more than 500,000 megawatt-hours of electricity. The demonstration plant has met its objectives. The project showed that more than 95 percent of sulfur dioxide pollutants could be removed inside the advanced boiler using the advanced combustion technology, giving future power plants an attractive alternative to expensive, add-on scrubber technology. In addition to its sulfur removal effectiveness, the plant`s sustained periods of steady-state operation boosted its availability significantly above design projections, heightening confidence that pressurized fluidized bed technology will be a reliable, baseload technology for future power plants. The technology also controlled the release of nitrogen oxides to levels well below the allowable limits set by federal air quality standards. It also produced a dry waste product that is much easier to handle than wastes from conventional power plants and will likely have commercial value when produced by future power plants.

  12. Practical applications of probabilistic structural reliability analyses to primary pressure systems of nuclear power plants

    Witt, F.J.

    1980-01-01

    Primary pressure systems of nuclear power plants are built to exacting codes and standards with provisions for inservice inspection and repair if necessary. Analyses and experiments have demonstrated by deterministic means that very large margins exist on safety impacting failures under normal operating and upset conditions. Probabilistic structural reliability analyses provide additional support that failures of significance are very, very remote. They may range in degree of sophistication from very simple calculations to very complex computer analyses involving highly developed mathematical techniques. The end result however should be consistent with the desired usage. In this paper a probabilistic structural reliability analysis is performed as a supplement to in-depth deterministic evaluations with the primary objective to demonstrate an acceptably low probability of failure for the conditions considered. (author)

  13. Vaccine Safety

    ... During Pregnancy Frequently Asked Questions about Vaccine Recalls Historical Vaccine Safety Concerns FAQs about GBS and Menactra ... CISA Resources for Healthcare Professionals Evaluation Current Studies Historical Background 2001-12 Publications Technical Reports Vaccine Safety ...

  14. SAFETY FIRST

    2007-01-01

    Ensuring safety while peacefully utilizing nuclear energy is a top priority for China A fter a recent earthquake in Japan caused radioactive leaks at a nuclear power plant in Tokyo, the safety of nuclear energy has again aroused public attention.

  15. Water Safety

    ... Staying Safe Videos for Educators Search English Español Water Safety KidsHealth / For Parents / Water Safety What's in ... remains your best measure of protection. Making Kids Water Wise It's important to teach your kids proper ...

  16. Kinesthetic Transverse Wave Demonstration

    Pantidos, Panagiotis; Patapis, Stamatis

    2005-09-01

    This is a variation on the String and Sticky Tape demonstration "The Wave Game," suggested by Ron Edge. A group of students stand side by side, each one holding a card chest high with both hands. The teacher cues the first student to begin raising and lowering his card. When he starts lowering his card, the next student begins to raise his. As succeeding students move their cards up and down, a wave such as that shown in the figure is produced. To facilitate the process, students' motions were synchronized with the ticks of a metronome (without such synchronization it was nearly impossible to generate a satisfactory wave). Our waves typically had a frequency of about 1 Hz and a wavelength of around 3 m. We videotaped the activity so that the students could analyze the motions. The (17-year-old) students had not received any prior instruction regarding wave motion and did not know beforehand the nature of the exercise they were about to carry out. During the activity they were asked what a transverse wave is. Most of them quickly realized, without teacher input, that while the wave propagated horizontally, the only motion of the transmitting medium (them) was vertical. They located the equilibrium points of the oscillations, the crests and troughs of the waves, and identified the wavelength. The teacher defined for them the period of the oscillations of the motion of a card to be the total time for one cycle. The students measured this time and then several asserted that it was the same as the wave period. Knowing the length of the waves and the number of waves per second, the next step can easily be to find the wave speed.

  17. [Safety culture: definition, models and design].

    Pfaff, Holger; Hammer, Antje; Ernstmann, Nicole; Kowalski, Christoph; Ommen, Oliver

    2009-01-01

    Safety culture is a multi-dimensional phenomenon. Safety culture of a healthcare organization is high if it has a common stock in knowledge, values and symbols in regard to patients' safety. The article intends to define safety culture in the first step and, in the second step, demonstrate the effects of safety culture. We present the model of safety behaviour and show how safety culture can affect behaviour and produce safe behaviour. In the third step we will look at the causes of safety culture and present the safety-culture-model. The main hypothesis of this model is that the safety culture of a healthcare organization strongly depends on its communication culture and its social capital. Finally, we will investigate how the safety culture of a healthcare organization can be improved. Based on the safety culture model six measures to improve safety culture will be presented.

  18. Methods and Effects of Safety Enhancement in Korean PSR

    Kim, Young Gab; Park, Jong Woon

    2009-01-01

    Periodic Safety Review (PSR) is a comprehensive study on a nuclear power plant safety, taking into account aspects such as operational history, ageing, safety analyses and advances in code and standards since the time of construction. In Korea, PSRs have been performed for 20 units and have been effectively used to obtain an overall view of actual plant safety to determine reasonable and practical modifications that should be made in order to obtain a higher level of safety approaching that of modern plants. Among many safety enhancements achieved from Korean PSRs, new safety analyses are the important methods to confirm plant safety by increasing safety margin for specific safety issues. Methods and effects of safety enhancements applied in Korean PSRs are reviewed in this paper in light of new safety analyses to obtain additional safety margins

  19. Laquinimod Safety Profile

    Sørensen, Per Soelberg; Comi, Giancarlo; Vollmer, Timothy L

    2017-01-01

    the safety profile of laquinimod versus placebo. Adverse events (AEs), laboratory value changes, and potential risks identified in preclinical studies were evaluated in participants in ALLEGRO and BRAVO treated with at least one dose of laquinimod or matching placebo (1:1 random assignment). RESULTS...... laquinimod studies demonstrate a safety profile comprising benign or manageable AEs and asymptomatic laboratory findings with a clear temporal pattern. Potential risks noted in preclinical studies were not observed....

  20. Objectives of safety evaluation

    Rosen, M.

    1980-01-01

    An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants. In view of this and the generally small regulatory staffs of importing coutnries, suggestions are given for measures which should be taken by various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. (orig.)