WorldWideScience

Sample records for safety decisions analysis

  1. Safety analysis in support of regulatory decision marking

    International Nuclear Information System (INIS)

    Pomier Baez, L.; Troncoso Fleitas, M.; Valhuerdi Debesa, C.; Valle Cepero, R.; Hernandez, J.L.

    1996-01-01

    Features of different safety analysis techniques by means of calculation thermohydraulic a probabilistic and severe accidents used in the safety assessment, as well as the development of these techniques in Cuba and their use in support of regulatory decision making are presented

  2. A safety decision analysis for Saudi Arabian nuclear research facility

    International Nuclear Information System (INIS)

    Abulfaraj, W.H.; Abdul-Fattah, A.F.

    1985-01-01

    Establishment of a nuclear research facility should be the first step in planning for introducing the nuclear energy to Saudi Arabia. The fuzzy set decision theory is selected among different decision theories to be applied for this analysis. Four research reactors from USA are selected for the present study. The IFDA computer code, based on the fuzzy set theory is applied. Results reveal that the FNR reactor is the best alternative for the case of Saudi Arabian nuclear research facility, and MITR is the second best. 17 refs

  3. Safety margins of operating reactors. Analysis of uncertainties and implications for decision making

    International Nuclear Information System (INIS)

    2003-01-01

    Maintaining safety in the design and operation of nuclear power plants (NPPs) is a very important task under the conditions of a challenging environment, affected by the deregulated electricity market and implementation of risk informed regulations. In Member States, advanced computer codes are widely used as safety analysis tools in the framework of licensing of new NPP projects, safety upgrading programmes of existing NPPs, periodic safety reviews, renewal of operating licences, use of the safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, for justification of lifetime extensions, development of new emergency operating procedures, analysis of operational events, and development of accident management programmes. The issue of inadequate quality of safety analysis is becoming important due to a general tendency to use advanced tools for better establishment and utilization of safety margins, while the existence of such margins assure that NPPs operate safely in all modes of operation and at all times. The most important safety margins relate to physical barriers against release of radioactive material, such as fuel matrix and fuel cladding, reactor coolant system boundary, and the containment. Typically, safety margins are determined with use of computational tools for safety analysis. Advanced best estimate computer codes are suggested e.g. in the IAEA Safety Guide on Safety Assessment and Verification for Nuclear Power Plants to be used for current safety analysis. Such computer codes require their careful application to avoid unjustified reduction in robustness of the reactor safety. The issue of uncertainties in safety analyses and their impact on evaluation of safety margins is addressed in a number of IAEA guidance documents, in particular in the Safety Report on Accident Analysis for Nuclear Power Plants. It is also discussed in various technical meetings and workshops devoted to this area. The

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. Adapting Cognitive Task Analysis to Investigate Clinical Decision Making and Medication Safety Incidents.

    Science.gov (United States)

    Russ, Alissa L; Militello, Laura G; Glassman, Peter A; Arthur, Karen J; Zillich, Alan J; Weiner, Michael

    2017-05-03

    Cognitive task analysis (CTA) can yield valuable insights into healthcare professionals' cognition and inform system design to promote safe, quality care. Our objective was to adapt CTA-the critical decision method, specifically-to investigate patient safety incidents, overcome barriers to implementing this method, and facilitate more widespread use of cognitive task analysis in healthcare. We adapted CTA to facilitate recruitment of healthcare professionals and developed a data collection tool to capture incidents as they occurred. We also leveraged the electronic health record (EHR) to expand data capture and used EHR-stimulated recall to aid reconstruction of safety incidents. We investigated 3 categories of medication-related incidents: adverse drug reactions, drug-drug interactions, and drug-disease interactions. Healthcare professionals submitted incidents, and a subset of incidents was selected for CTA. We analyzed several outcomes to characterize incident capture and completed CTA interviews. We captured 101 incidents. Eighty incidents (79%) met eligibility criteria. We completed 60 CTA interviews, 20 for each incident category. Capturing incidents before interviews allowed us to shorten the interview duration and reduced reliance on healthcare professionals' recall. Incorporating the EHR into CTA enriched data collection. The adapted CTA technique was successful in capturing specific categories of safety incidents. Our approach may be especially useful for investigating safety incidents that healthcare professionals "fix and forget." Our innovations to CTA are expected to expand the application of this method in healthcare and inform a wide range of studies on clinical decision making and patient safety.

  6. Does a reactor need a safety backfit. Case study on communicating decision and risk analysis information to managers

    Energy Technology Data Exchange (ETDEWEB)

    Brown, R.V.; Ulvila, J.W.

    1988-06-01

    An approach to communicating decision and risk analysis findings to managers is illustrated in a real case context. This article consists essentially of a report prepared for senior managers of the Nuclear Regulatory Commission to help them make a reactor safety decision. It illustrates the communication of decision analysis findings relating to technical risks, costs, and benefits in support of a major risk management decision: whether or not to require a safety backfit. Its focus is on the needs of decision makers, and it introduces some novel communication devices.

  7. Frame-based safety analysis approach for decision-based errors

    International Nuclear Information System (INIS)

    Fan, Chin-Feng; Yihb, Swu

    1997-01-01

    A frame-based approach is proposed to analyze decision-based errors made by automatic controllers or human operators due to erroneous reference frames. An integrated framework, Two Frame Model (TFM), is first proposed to model the dynamic interaction between the physical process and the decision-making process. Two important issues, consistency and competing processes, are raised. Consistency between the physical and logic frames makes a TFM-based system work properly. Loss of consistency refers to the failure mode that the logic frame does not accurately reflect the state of the controlled processes. Once such failure occurs, hazards may arise. Among potential hazards, the competing effect between the controller and the controlled process is the most severe one, which may jeopardize a defense-in-depth design. When the logic and physical frames are inconsistent, conventional safety analysis techniques are inadequate. We propose Frame-based Fault Tree; Analysis (FFTA) and Frame-based Event Tree Analysis (FETA) under TFM to deduce the context for decision errors and to separately generate the evolution of the logical frame as opposed to that of the physical frame. This multi-dimensional analysis approach, different from the conventional correctness-centred approach, provides a panoramic view in scenario generation. Case studies using the proposed techniques are also given to demonstrate their usage and feasibility

  8. Decision support systems and expert systems for risk and safety analysis

    International Nuclear Information System (INIS)

    Baybutt, P.

    1986-01-01

    During the last 1-2 years, rapid developments have occurred in the development of decision support systems and expert systems to aid in decision making related to risk and safety of industrial plants. These activities are most noteworthy in the nuclear industry where numerous systems are under development with implementation often being made on personal computers. An overview of some of these developments is provided, and an example of one recently developed decision support system is given. This example deals with CADET, a system developed to aid the U.S. Nuclear Regulatory Commission in making decisions related to the topical issue of source terms resulting from degraded core accidents in light water reactors. The paper concludes with some comments on the likely directions of future developments in decision support systems and expert systems to aid in the management of risk and safety in industrial plants. (author)

  9. Markov Modeling with Soft Aggregation for Safety and Decision Analysis; TOPICAL

    International Nuclear Information System (INIS)

    COOPER, J. ARLIN

    1999-01-01

    The methodology in this report improves on some of the limitations of many conventional safety assessment and decision analysis methods. A top-down mathematical approach is developed for decomposing systems and for expressing imprecise individual metrics as possibilistic or fuzzy numbers. A ''Markov-like'' model is developed that facilitates combining (aggregating) inputs into overall metrics and decision aids, also portraying the inherent uncertainty. A major goal of Markov modeling is to help convey the top-down system perspective. One of the constituent methodologies allows metrics to be weighted according to significance of the attribute and aggregated nonlinearly as to contribution. This aggregation is performed using exponential combination of the metrics, since the accumulating effect of such factors responds less and less to additional factors. This is termed ''soft'' mathematical aggregation. Dependence among the contributing factors is accounted for by incorporating subjective metrics on ''overlap'' of the factors as well as by correspondingly reducing the overall contribution of these combinations to the overall aggregation. Decisions corresponding to the meaningfulness of the results are facilitated in several ways. First, the results are compared to a soft threshold provided by a sigmoid function. Second, information is provided on input ''Importance'' and ''Sensitivity,'' in order to know where to place emphasis on considering new controls that may be necessary. Third, trends in inputs and outputs are tracked in order to obtain significant information% including cyclic information for the decision process. A practical example from the air transportation industry is used to demonstrate application of the methodology. Illustrations are given for developing a structure (along with recommended inputs and weights) for air transportation oversight at three different levels, for developing and using cycle information, for developing Importance and

  10. Decision analysis multicriteria analysis

    International Nuclear Information System (INIS)

    Lombard, J.

    1986-09-01

    The ALARA procedure covers a wide range of decisions from the simplest to the most complex one. For the simplest one the engineering judgement is generally enough and the use of a decision aiding technique is therefore not necessary. For some decisions the comparison of the available protection option may be performed from two or a few criteria (or attributes) (protection cost, collective dose,...) and the use of rather simple decision aiding techniques, like the Cost Effectiveness Analysis or the Cost Benefit Analysis, is quite enough. For the more complex decisions, involving numerous criteria or for decisions involving large uncertainties or qualitative judgement the use of these techniques, even the extended cost benefit analysis, is not recommended and appropriate techniques like multi-attribute decision aiding techniques are more relevant. There is a lot of such particular techniques and it is not possible to present all of them. Therefore only two broad categories of multi-attribute decision aiding techniques will be presented here: decision analysis and the outranking analysis

  11. Environmental, health, and safety decision making for naturally occurring radioactive materials in producing operations using pathway exposure analysis

    International Nuclear Information System (INIS)

    Miller, H.T.; Cook, L.M.

    1991-01-01

    A number of health and safety issues have arisen because of the occurrence of NORM, naturally occurring radioactive materials of the 226 radium and 228 radium decay chains, in production operations. Issues such as risk to workers or the general public, disposal of contaminated production fluids, disposal of NORM removed in cleaning equipment and tubing, and procedures to follow in well rework, equipment decontamination and other types of maintenance must be addressed. This paper describes the application of a procedural aid to decision making known as pathway exposure analysis to these issues. The procedure examines the radiation exposure of individuals and population groups by calculating the dose from each exposure route and pathway. The sum of these is used to calculate the overall risk to the individual or the group. This method can be used to examine management and procedural options to identify the option offering the smallest risk. Risk information coupled with cost estimates then permits management maximum utilization of its available resources

  12. TWRS Final Safety Analysis Report (FSAR) integrated control decision team (ICDT) meetings January 22 - 31,1997

    International Nuclear Information System (INIS)

    Saladin, V.L.

    1997-01-01

    U.S. Department of Energy (DOE), Richland Operations Office (RL) letter 97-MSD-163 dated January 15, 1997, directed the Project Hanford Management Contractor (Contractor), Fluor Daniel Hanford, inc., to form a joint RL-Contractor Integrated Control Decision Team (ICDT) to evaluate the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR) accident scenarios that were identified to be above the risk evaluation guidelines (radiological and/or toxicological) defined by the April 8, 1996, letter from J. Kinzer, RL-TWRS (96-MSO-069) to Dr. A. L. Trego, Westinghouse Hanford Company. The ICDT evaluated six postulated accidents from the draft FSAR which had analyzed consequences above the DOE directed risk evaluation guidelines after controls were applied. The accidents were: (1) Organic Solvent Fires; (2) Organic Salt-Nitrate Fire; (3) Spray Leak; (4) Flammable Gas; (5) Steam Intrusion; and (6) Seismic Event. Five of the postulated accidents exceed radiological risk guidelines. Although the postulated steam intrusion accident does not exceed the radiological risk guidelines, it was considered in the ICDT evaluation because its calculated consequences exceed toxicological risk evaluation guidelines. Figure 1 delineates the mitigated and unmitigated risk evaluations performed for the FSAR

  13. Safety analysis fundamentals

    International Nuclear Information System (INIS)

    Wright, A.C.D.

    2002-01-01

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

  14. Safety assessment as basis for the decision making process

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. Probability and uncertainty in nuclear safety decisions

    International Nuclear Information System (INIS)

    Pate-Cornell, M.E.

    1986-01-01

    In this paper, we examine some problems posed by the use of probabilities in Nuclear Safety decisions. We discuss some of the theoretical difficulties due to the collective nature of regulatory decisions, and, in particular, the calibration and the aggregation of risk information (e.g., experts opinions). We argue that, if one chooses numerical safety goals as a regulatory basis, one can reduce the constraints to an individual safety goal and a cost-benefit criterion. We show the relevance of risk uncertainties in this kind of regulatory framework. We conclude that, whereas expected values of future failure frequencies are adequate to show compliance with economic constraints, the use of a fractile (e.g., 95%) to be specified by the regulatory agency is justified to treat hazard uncertainties for the individual safety goal. (orig.)

  16. Regulator Loss Functions and Hierarchical Modeling for Safety Decision Making.

    Science.gov (United States)

    Hatfield, Laura A; Baugh, Christine M; Azzone, Vanessa; Normand, Sharon-Lise T

    2017-07-01

    Regulators must act to protect the public when evidence indicates safety problems with medical devices. This requires complex tradeoffs among risks and benefits, which conventional safety surveillance methods do not incorporate. To combine explicit regulator loss functions with statistical evidence on medical device safety signals to improve decision making. In the Hospital Cost and Utilization Project National Inpatient Sample, we select pediatric inpatient admissions and identify adverse medical device events (AMDEs). We fit hierarchical Bayesian models to the annual hospital-level AMDE rates, accounting for patient and hospital characteristics. These models produce expected AMDE rates (a safety target), against which we compare the observed rates in a test year to compute a safety signal. We specify a set of loss functions that quantify the costs and benefits of each action as a function of the safety signal. We integrate the loss functions over the posterior distribution of the safety signal to obtain the posterior (Bayes) risk; the preferred action has the smallest Bayes risk. Using simulation and an analysis of AMDE data, we compare our minimum-risk decisions to a conventional Z score approach for classifying safety signals. The 2 rules produced different actions for nearly half of hospitals (45%). In the simulation, decisions that minimize Bayes risk outperform Z score-based decisions, even when the loss functions or hierarchical models are misspecified. Our method is sensitive to the choice of loss functions; eliciting quantitative inputs to the loss functions from regulators is challenging. A decision-theoretic approach to acting on safety signals is potentially promising but requires careful specification of loss functions in consultation with subject matter experts.

  17. Review of cause-based decision tree approach for the development of domestic standard human reliability analysis procedure in low power/shutdown operation probabilistic safety assessment

    International Nuclear Information System (INIS)

    Kang, D. I.; Jung, W. D.

    2003-01-01

    We review the Cause-Based Decision Tree (CBDT) approach to decide whether we incorporate it or not for the development of domestic standard Human Reliability Analysis (HRA) procedure in low power/shutdown operation Probabilistic Safety Assessment (PSA). In this paper, we introduce the cause based decision tree approach, quantify human errors using it, and identify merits and demerits of it in comparision with previously used THERP. The review results show that it is difficult to incorporate the CBDT method for the development of domestic standard HRA procedure in low power/shutdown PSA because the CBDT method need for the subjective judgment of HRA analyst like as THERP. However, it is expected that the incorporation of the CBDT method into the development of domestic standard HRA procedure only for the comparision of quantitative HRA results will relieve the burden of development of detailed HRA procedure and will help maintain consistent quantitative HRA results

  18. Plutonium-238 Decision Analysis

    International Nuclear Information System (INIS)

    Brown, Mike; Lechel, David J.; Leigh, C.D.

    1999-01-01

    Five transuranic (TRU) waste sites in the Department of Energy (DOE) complex, collectively, have more than 2,100 cubic meters of Plutonium-238 (Pu-238) TRU waste that exceed the wattage restrictions of the Transuranic Package Transporter-II (TRUPACT-11). The Waste Isolation Pilot Plant (WIPP) is being developed by the DOE as a repository for TRU waste. With the Waste Isolation Pilot Plant (WIPP) opening in 1999, these sites are faced with a need to develop waste management practices that will enable the transportation of Pu-238 TRU waste to WIPP for disposal. This paper describes a decision analysis that provided a logical framework for addressing the Pu-238 TRU waste issue. The insights that can be gained by performing a formalized decision analysis are multifold. First and foremost, the very process. of formulating a decision tree forces the decision maker into structured, logical thinking where alternatives can be evaluated one against the other using a uniform set of criteria. In the process of developing the decision tree for transportation of Pu-238 TRU waste, several alternatives were eliminated and the logical order for decision making was discovered. Moreover, the key areas of uncertainty for proposed alternatives were identified and quantified. The decision analysis showed that the DOE can employ a combination approach where they will (1) use headspace gas analyses to show that a fraction of the Pu-238 TRU waste drums are no longer generating hydrogen gas and can be shipped to WIPP ''as-is'', (2) use drums and bags with advanced filter systems to repackage Pu-238 TRU waste drums that are still generating hydrogen, and (3) add hydrogen getter materials to the inner containment vessel of the TRUPACT-11to relieve the build-up of hydrogen gas during transportation of the Pu-238 TRU waste drums

  19. Road Safety Data, Collection, Transfer and Analysis DaCoTa. Workpackage 4, Decision Support: Deliverable 4.9: Developing a road safety index.

    NARCIS (Netherlands)

    Bax, C.A. Wesemann, P. Gitelman, V. Shen, Y. Goldenbeld, C. Hermans, E. Doveh, E. Hakkert, S. Wegman, F.C.M. & Aarts, L.T.

    2015-01-01

    Road safety is a major social aim. The countries that perform best in road safety base their most effective policies on an evidence-based, scientific approach. Countries may learn to improve road safety from their own experiences but also from systematic comparison with other countries. This study

  20. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

    Delage, M.; Giroux, C.

    1990-12-01

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

  1. Use of decision analytic methods in nuclear safety. An international survey

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J.; Pulkkinen, U. [VTT Automation, Espoo (Finland). Industrial Automation

    1996-12-01

    This report reviews applications of formal decision analysis methods in resolving nuclear safety related issues. The review is based on selected published reports and a questionnaire sent to the members of the Principal Working Group 5 on risk analysis (PWG5) of OECD/NEA/CSNI. In the report, decision analysis methodology is shortly described. The applications discussed in this review are related to probabilistic safety goals of safety criteria, operational safety management, nuclear waste management and emergency management. The experiences from the application decision analysis methodology have been mainly positive. The advantages provided by the decision analytical thinking are the structured view over the problem under consideration and the explicit statements on uncertainties, values and preferences. The decision analysis methodology is rather mature to be applied in solution of nuclear safety issues. Although the applications have been mainly research oriented, it can be expected that the practical use of the methodology shall be more common in future. (orig.) (27 refs.).

  2. Use of decision analytic methods in nuclear safety. An international survey

    International Nuclear Information System (INIS)

    Holmberg, J.; Pulkkinen, U.

    1996-12-01

    This report reviews applications of formal decision analysis methods in resolving nuclear safety related issues. The review is based on selected published reports and a questionnaire sent to the members of the Principal Working Group 5 on risk analysis (PWG5) of OECD/NEA/CSNI. In the report, decision analysis methodology is shortly described. The applications discussed in this review are related to probabilistic safety goals of safety criteria, operational safety management, nuclear waste management and emergency management. The experiences from the application decision analysis methodology have been mainly positive. The advantages provided by the decision analytical thinking are the structured view over the problem under consideration and the explicit statements on uncertainties, values and preferences. The decision analysis methodology is rather mature to be applied in solution of nuclear safety issues. Although the applications have been mainly research oriented, it can be expected that the practical use of the methodology shall be more common in future. (orig.) (27 refs.)

  3. The use of efficiency assessment tools : solutions to barriers : Workpackage 3 of the European research project ROSEBUD (Road Safety and Environmental Cost-Benefit and Cost-Effectiveness Analysis for Use in Decision-making).

    NARCIS (Netherlands)

    Hakkert, A.S. & Wesemann, P. (eds.)

    2005-01-01

    In road safety, as in most other fields, efficiency is an important criterion in political and professional decision making. Efficiency Assessment Tools (EATs) like Cost Benefit Analysis and Cost Effectiveness Analysis are available to help choose the policy which gives the highest return on

  4. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    2000-01-01

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

  5. K Basin safety analysis

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

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

  7. Advancing Alternative Analysis: Integration of Decision Science.

    Science.gov (United States)

    Malloy, Timothy F; Zaunbrecher, Virginia M; Batteate, Christina M; Blake, Ann; Carroll, William F; Corbett, Charles J; Hansen, Steffen Foss; Lempert, Robert J; Linkov, Igor; McFadden, Roger; Moran, Kelly D; Olivetti, Elsa; Ostrom, Nancy K; Romero, Michelle; Schoenung, Julie M; Seager, Thomas P; Sinsheimer, Peter; Thayer, Kristina A

    2017-06-13

    Decision analysis-a systematic approach to solving complex problems-offers tools and frameworks to support decision making that are increasingly being applied to environmental challenges. Alternatives analysis is a method used in regulation and product design to identify, compare, and evaluate the safety and viability of potential substitutes for hazardous chemicals. We assessed whether decision science may assist the alternatives analysis decision maker in comparing alternatives across a range of metrics. A workshop was convened that included representatives from government, academia, business, and civil society and included experts in toxicology, decision science, alternatives assessment, engineering, and law and policy. Participants were divided into two groups and were prompted with targeted questions. Throughout the workshop, the groups periodically came together in plenary sessions to reflect on other groups' findings. We concluded that the further incorporation of decision science into alternatives analysis would advance the ability of companies and regulators to select alternatives to harmful ingredients and would also advance the science of decision analysis. We advance four recommendations: a ) engaging the systematic development and evaluation of decision approaches and tools; b ) using case studies to advance the integration of decision analysis into alternatives analysis; c ) supporting transdisciplinary research; and d ) supporting education and outreach efforts. https://doi.org/10.1289/EHP483.

  8. Probabilistic safety assessment as a standpoint for decision making

    International Nuclear Information System (INIS)

    Cepin, M.

    2001-01-01

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

  9. Safety analysis for 'Fugen'

    International Nuclear Information System (INIS)

    1997-10-01

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

  10. Procurement in the Nuclear Industry, Quality, Safety and Decision Making

    International Nuclear Information System (INIS)

    Jakobsson, Marianne; Svenson, Ola; Salo, Ilkka

    2010-03-01

    The major purpose of the present study is partly to map and partly to make an analysis of the decision processes in the procurement routines in the nuclear industry in order to provide a basis for: 1. further development of safety inspections about procurements for Swedish Radiation Safety Authority 2. improvements of safety management in connection with procurement within a nuclear-power plant, 3 improvements of procurement routines in general in a nuclear power plant. The procurement processes at a nuclear power plant were analyzed from a decision theoretic perspective. Key staff at the plant was interviewed and written instructions as well as digitalized processes were used in the analysis. The results illustrate the most important moments during the procurement process with descriptions from interviews and documents. The staff at the nuclear power plant used a multi-attribute utility decision theory MAUT-inspired model in evaluation of alternatives and both compensatory (in which negative aspects can be compensated by positive aspects) and non-compensatory (in which certain 'pass' levels of attributes have to be exceeded for a choice) decision rules were used in the procurement process. Not surprising, nuclear safety was evaluated in a non-compensatory manner following regulatory criteria while costs were evaluated in trade-off compensatory rules, which means that a weakness in one consideration might be compensated by strength in another consideration. Thus, nuclear safety above the regulator's and law requirements are not integrated in a compensatory manner when procurement alternatives are evaluated. The nuclear plant assessed an organization's safety culture at an early stage of the purchasing process. A successful and a less successful procurement case were reported with the lessons learned from them. We find that the existing written instructions for purchase were well elaborated and adequate. There is a lack of personal resources when procurement teams

  11. Decision Analysis Technique

    Directory of Open Access Journals (Sweden)

    Hammad Dabo Baba

    2014-01-01

    Full Text Available One of the most significant step in building structure maintenance decision is the physical inspection of the facility to be maintained. The physical inspection involved cursory assessment of the structure and ratings of the identified defects based on expert evaluation. The objective of this paper is to describe present a novel approach to prioritizing the criticality of physical defects in a residential building system using multi criteria decision analysis approach. A residential building constructed in 1985 was considered in this study. Four criteria which includes; Physical Condition of the building system (PC, Effect on Asset (EA, effect on Occupants (EO and Maintenance Cost (MC are considered in the inspection. The building was divided in to nine systems regarded as alternatives. Expert's choice software was used in comparing the importance of the criteria against the main objective, whereas structured Proforma was used in quantifying the defects observed on all building systems against each criteria. The defects severity score of each building system was identified and later multiplied by the weight of the criteria and final hierarchy was derived. The final ranking indicates that, electrical system was considered the most critical system with a risk value of 0.134 while ceiling system scored the lowest risk value of 0.066. The technique is often used in prioritizing mechanical equipment for maintenance planning. However, result of this study indicates that the technique could be used in prioritizing building systems for maintenance planning

  12. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

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

  13. Safety margins in deterministic safety analysis

    International Nuclear Information System (INIS)

    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)

  14. Decision making under uncertainty: An investigation into the application of formal decision-making methods to safety issue decisions

    International Nuclear Information System (INIS)

    Bohn, M.P.

    1992-12-01

    As part of the NRC-sponsored program to study the implications of Generic Issue 57, ''Effects of Fire Protection System Actuation on Safety-Related Equipment,'' a subtask was performed to evaluate the applicability of formal decision analysis methods to generic issues cost/benefit-type decisions and to apply these methods to the GI-57 results. In this report, the numerical results obtained from the analysis of three plants (two PWRs and one BWR) as developed in the technical resolution program for GI-57 were studied. For each plant, these results included a calculation of the person-REM averted due to various accident scenarios and various proposed modifications to mitigate the accident scenarios identified. These results were recomputed to break out the benefit in terms of contributions due to random event scenarios, fire event scenarios, and seismic event scenarios. Furthermore, the benefits associated with risk (in terms of person-REM) averted from earthquakes at three different seismic ground motion levels were separately considered. Given this data, formal decision methodologies involving decision trees, value functions, and utility functions were applied to this basic data. It is shown that the formal decision methodology can be applied at several different levels. Examples are given in which the decision between several retrofits is changed from that resulting from a simple cost/benefit-ratio criterion by virtue of the decision-makinger's expressed (and assumed) preferences

  15. Ignalina Safety Analysis Group

    International Nuclear Information System (INIS)

    Ushpuras, E.

    1995-01-01

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

  16. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

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

  17. Subseabed disposal safety analysis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. Computer aided safety analysis

    International Nuclear Information System (INIS)

    1988-05-01

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

  19. Safety-related decision making at a nuclear power plant

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1998-01-01

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

  20. Reactor safety training for decision making

    International Nuclear Information System (INIS)

    Scott, C.K.

    2003-01-01

    The purpose of this paper is to describe an approach to reactor safety training for technical staff working at an operating station. The concept being developed is that, when the engineer becomes a registered professional engineer, they have sufficient reactor safety knowledge to perform independent technical work without compromising the safety of the plant. This goal would be achieved with a focused training program while working as an engineer-in-training (four years in NB). (author)

  1. Software safety hazard analysis

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1996-02-01

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

  2. Probabilistic safety analysis procedures guide

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Buslik, A.J.

    1984-01-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of tissues affecting reactor safety. This guide addresses the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant and from loss of offsite electric power. The scope includes analyses of problem-solving (cognitive) human errors, a determination of importance of the various core damage accident sequences, and an explicit treatment and display of uncertainties for the key accident sequences. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance) and the risk associated with external accident initiators, as consensus is developed regarding suitable methodologies in these areas. This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are essential for regulatory decision making. Methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study

  3. Risk analysis and safety rationale

    International Nuclear Information System (INIS)

    Bengtsson, G.

    1989-01-01

    Decision making with respect to safety is becoming more and more complex. The risk involved must be taken into account together with numerous other factors such as the benefits, the uncertainties and the public perception. Can the decision maker be aided by some kind of system, general rules of thumb, or broader perspective on similar decisions? This question has been addressed in a joint Nordic project relating to nuclear power. Modern techniques for risk assessment and management have been studied, and parallels drawn to such areas as offshore safety and management of toxic chemicals in the environment. The report summarises the finding of 5 major technical reports which have been published in the NORD-series. The topics includes developments, uncertainties and limitations in probabilistic safety assessments, negligible risks, risk-cost trade-offs, optimisation of nuclear safety and radiation protection, and the role of risks in the decision making process. (author) 84 refs

  4. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

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

  5. Towards an integrated approach in supporting microbiological food safety decisions

    NARCIS (Netherlands)

    Havelaar, A.H.; Bräunig, J.; Christiansen, K.; Cornu, M.; Hald, T.; Mangen, M.J.J.; Molbak, K.; Pielaat, A.; Snary, E.; Pelt, van W.; Velthuis, A.G.J.; Wahlström, H.

    2007-01-01

    Decisions on food safety involve consideration of a wide range of concerns including the public health impact of foodborne illness, the economic importance of the agricultural sector and the food industry, and the effectiveness and efficiency of interventions. To support such decisions, we propose

  6. Fire safety analysis: methodology

    International Nuclear Information System (INIS)

    Kazarians, M.

    1998-01-01

    From a review of the fires that have occurred in nuclear power plants and the results of fire risk studies that have been completed over the last 17 years, we can conclude that internal fires in nuclear power plants can be an important contributor to plant risk. Methods and data are available to quantify the fire risk. These methods and data have been subjected to a series of reviews and detailed scrutiny and have been applied to a large number of plants. There is no doubt that we do not know everything about fire and its impact on a nuclear power plants. However, this lack of knowledge or uncertainty can be quantified and can be used in the decision making process. In other words, the methods entail uncertainties and limitations that are not insurmountable and there is little or no basis for the results of a fire risk analysis fail to support a decision process

  7. Lone ranger decision making versus consensus decision making: Descriptive analysis

    OpenAIRE

    Maite Sara Mashego

    2015-01-01

    Consensus decision making, concerns group members make decisions together with the requirement of reaching a consensus that is all members abiding by the decision outcome. Lone ranging worked for sometime in a autocratic environment. Researchers are now pointing to consensus decision-making in organizations bringing dividend to many organizations. This article used a descriptive analysis to compare the goodness of consensus decision making and making lone ranging decision management. This art...

  8. Development of a safety decision-making scenario to measure worker safety in agriculture.

    Science.gov (United States)

    Mosher, G A; Keren, N; Freeman, S A; Hurburgh, C R

    2014-04-01

    Human factors play an important role in the management of occupational safety, especially in high-hazard workplaces such as commercial grain-handling facilities. Employee decision-making patterns represent an essential component of the safety system within a work environment. This research describes the process used to create a safety decision-making scenario to measure the process that grain-handling employees used to make choices in a safety-related work task. A sample of 160 employees completed safety decision-making simulations based on a hypothetical but realistic scenario in a grain-handling environment. Their choices and the information they used to make their choices were recorded. Although the employees emphasized safety information in their decision-making process, not all of their choices were safe choices. Factors influencing their choices are discussed, and implications for industry, management, and workers are shared.

  9. Ending on a positive: Examining the role of safety leadership decisions, behaviours and actions in a safety critical situation.

    Science.gov (United States)

    Donovan, Sarah-Louise; Salmon, Paul M; Horberry, Timothy; Lenné, Michael G

    2018-01-01

    Safety leadership is an important factor in supporting safe performance in the workplace. The present case study examined the role of safety leadership during the Bingham Canyon Mine high-wall failure, a significant mining incident in which no fatalities or injuries were incurred. The Critical Decision Method (CDM) was used in conjunction with a self-reporting approach to examine safety leadership in terms of decisions, behaviours and actions that contributed to the incidents' safe outcome. Mapping the analysis onto Rasmussen's Risk Management Framework (Rasmussen, 1997), the findings demonstrate clear links between safety leadership decisions, and emergent behaviours and actions across the work system. Communication and engagement based decisions featured most prominently, and were linked to different leadership practices across the work system. Further, a core sub-set of CDM decision elements were linked to the open flow and exchange of information across the work system, which was critical to supporting the safe outcome. The findings provide practical implications for the development of safety leadership capability to support safety within the mining industry. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Importance of Decision Support Systems About Food Safety in Raw Milk Production

    Directory of Open Access Journals (Sweden)

    Ecem Akan

    2015-12-01

    Full Text Available In raw milk production decision support systems for control of food safety hazards has not been developed but main points of this system are available. The decision support systems’ elements include data identification at critical points in the milk supply chain, an information management system and data exchange. Decision supports systems has been developed on the basis of these elements. In dairy sector decision support systems are significant for controlling of food safety hazards and preferred by producers. When these systems are implemented in the milk supply chain, it can be prevented unnecessary sampling and analysis. In this article it will be underlined effects of decision support system elements on food safety of raw milk.

  11. Preclosure Safety Analysis Guide

    International Nuclear Information System (INIS)

    D.D. Orvis

    2003-01-01

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

  12. Comparison of safety measures with a multicriteria decision aiding technique

    International Nuclear Information System (INIS)

    Lombard, J.

    1985-01-01

    Attributes such as political, social and psychological factors have to be taken into account for the decision-making process. Multiattribute decision-aiding techniques are used to cope with this multidimensionality of the risk management process. A simple example will be given to illustrate how such method can be helpful for the selection of proper safety measures in a rational way. (orig./HP) [de

  13. Approximate reasoning in decision analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, M M; Sanchez, E

    1982-01-01

    The volume aims to incorporate the recent advances in both theory and applications. It contains 44 articles by 74 contributors from 17 different countries. The topics considered include: membership functions; composite fuzzy relations; fuzzy logic and inference; classifications and similarity measures; expert systems and medical diagnosis; psychological measurements and human behaviour; approximate reasoning and decision analysis; and fuzzy clustering algorithms.

  14. Deep Borehole Disposal Safety Analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2016-10-01

    This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

  15. Multicriteria decision analysis: Overview and implications for environmental decision making

    Science.gov (United States)

    Hermans, Caroline M.; Erickson, Jon D.; Erickson, Jon D.; Messner, Frank; Ring, Irene

    2007-01-01

    Environmental decision making involving multiple stakeholders can benefit from the use of a formal process to structure stakeholder interactions, leading to more successful outcomes than traditional discursive decision processes. There are many tools available to handle complex decision making. Here we illustrate the use of a multicriteria decision analysis (MCDA) outranking tool (PROMETHEE) to facilitate decision making at the watershed scale, involving multiple stakeholders, multiple criteria, and multiple objectives. We compare various MCDA methods and their theoretical underpinnings, examining methods that most realistically model complex decision problems in ways that are understandable and transparent to stakeholders.

  16. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  17. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

    Yu, J.; Kim, J. Y.

    2006-06-01

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

  18. Safety analysis procedures for PHWR

    International Nuclear Information System (INIS)

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

    2004-03-01

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

  19. Risk concepts in UK nuclear safety decision-making

    International Nuclear Information System (INIS)

    Brighton, P.W.M.

    2001-01-01

    This paper discusses the concept of risk as understood in the UK, with particular reference to the use of probabilistic safety assessment (PSA) in nuclear safety decision making. The way 'risk' appears in UK fundamental legislation means that the concept cannot be limited to evaluation of numerical probabilities of physical harm. Rather the focus is on doing all that is reasonably practicable to reduce risks: this entails applying relevant good practice and then seeking further safety measures until the money, time and trouble required are grossly disproportionate to the residual risk. PSA is used to inform rather than dictate such decisions. This approach is reinforced by considering how far any practical PSA can be said to measure risk. The behaviour of complex socio-technical systems such as nuclear power stations does not meet the conditions under which probability theory can be applied in an absolutely objective statistical sense. Risk is not an intrinsic real property of such systems. Rather PSA is a synthesis of data and subjective expert judgements, dependent on the extent of detailed knowledge of the plant. There are many other aspects of engineering judgement involved in safety decisions which cannot be so captured. (author)

  20. Risk analysis as a decision tool

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Chakraborty, S.

    1985-01-01

    From 1983 - 1985 a lecture series entitled ''Risk-benefit analysis'' was held at the Swiss Federal Institute of Technology (ETH), Zurich, in cooperation with the Central Department for the Safety of Nuclear Installations of the Swiss Federal Agency of Energy Economy. In that setting the value of risk-oriented evaluation models as a decision tool in safety questions was discussed on a broad basis. Experts of international reputation from the Federal Republic of Germany, France, Canada, the United States and Switzerland have contributed to report in this joint volume on the uses of such models. Following an introductory synopsis on risk analysis and risk assessment the book deals with practical examples in the fields of medicine, nuclear power, chemistry, transport and civil engineering. Particular attention is paid to the dialogue between analysts and decision makers taking into account the economic-technical aspects and social values. The recent chemical disaster in the Indian city of Bhopal again signals the necessity of such analyses. All the lectures were recorded individually. (orig./HP) [de

  1. System safety engineering analysis handbook

    Science.gov (United States)

    Ijams, T. E.

    1972-01-01

    The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

  2. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

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

  3. Risk perception, safety goals and regulatory decision-making

    International Nuclear Information System (INIS)

    Hoegberg, Lars

    1998-01-01

    Deciding on 'how safe is safe enough?' includes value judgements with implications of an ethical and political nature. As regulators are accountable to governments, parliaments and the general public, regulatory decision-making should be characterized by transparency with respect to how such value judgements are reflected in risk assessments and regulatory decisions. Some approaches in this respect are discussed in the paper, based on more than fifteen years of experience in nuclear regulatory decision-making. Issues discussed include: (1) risk profiles and safety goals associated with severe reactor accidents--individual health risks, societal risks and risk of losing investments; (2) risk profile-based licensing of the Swedish SFR final disposal facility for low and intermediate level radioactive waste

  4. Employee direct participation in organisational decisions and workplace safety.

    Science.gov (United States)

    Widerszal-Bazyl, Maria; Warszewska-Makuch, Magdalena

    2008-01-01

    Managers from 192 companies filled out the Employee Direct Participation in Organisational Change questionnaire measuring employees' direct participation (DP) in organisational decisions. Four main forms of DP were identified: individual and group consultations, and individual and group delegation. Workplace safety was measured with the number of accidents, the number of employees working in hazardous conditions, accident absenteeism and sickness absence. Results showed that the 2 latter indicators were significantly related to some parameters of DP. Thus, companies that used face-to-face individual consultation had lower accident absenteeism than ones that did not. The same effect was true for group consultation with temporary groups, and individual and group delegation. Workplaces with high scores for scope for group consultation had lower accident absenteeism, and those with high scores for scope for group delegation had lower sickness absence. It was concluded that employee DP had a positive influence on workplace safety, even if involvement was not directly related to safety.

  5. Safety analysis reports - new strategies

    International Nuclear Information System (INIS)

    Booth, J.A.

    1994-01-01

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

  6. Selection of tolerable risk criteria for dam safety decision making

    International Nuclear Information System (INIS)

    Nielsen, N.M.; Hartford, D.N.D.; MacDonald, T.F.

    1994-01-01

    Risk assessment has received increasing attention in recent years as a means of aiding decision making on dams by providing systematic and rational methods for dealing with risk and uncertainty. Risk assessment is controversial and decisions affecting risk to life are the most controversial. Tolerable criteria, based on the risks that society is prepared to accept in order to avoid excessive costs, set bounds within which risk-based decisions may be made. The components of risk associated with dam safety are addressed on an individual basis and criteria established for each component, thereby permitting flexibility in the balance between component risk and avoiding the problems of placing a monetary value on life. The guiding principle of individual risk is that dams do not impose intolerable risks on any individual. A risk to life of 1 in 10 4 per annum is generally considered the maximum tolerable risk. When considering societal risk, the safety of a dam should be proportional to the consequences of its failure. Risks of financial losses beyond the corporation's ability to finance should be so low as to be considered negligible. 17 refs., 3 figs

  7. IAEA Review for Gap Analysis of Safety Analysis Capability

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  8. Applied decision analysis and risk evaluation

    International Nuclear Information System (INIS)

    Ferse, W.; Kruber, S.

    1995-01-01

    During 1994 the workgroup 'Applied Decision Analysis and Risk Evaluation; continued the work on the knowledge based decision support system XUMA-GEFA for the evaluation of the hazard potential of contaminated sites. Additionally a new research direction was started which aims at the support of a later stage of the treatment of contaminated sites: The clean-up decision. For the support of decisions arising at this stage, the methods of decision analysis will be used. Computational aids for evaluation and decision support were implemented and a case study at a waste disposal site in Saxony which turns out to be a danger for the surrounding groundwater ressource was initiated. (orig.)

  9. Radwaste Disposal Safety Analysis

    International Nuclear Information System (INIS)

    Hwang, Yong Soo; Kang, C. H.; Lee, Y. M.; Lee, S. H.; Jeong, J. T.; Choi, J. W.; Park, S. W.; Lee, H. S.; Kim, J. H.; Jeong, M. S.

    2010-02-01

    For the purpose of evaluating annual individual doses from a potential repository disposing of radioactive wastes from the operation of the prospective advanced nuclear fuel cycle facilities in Korea, the new safety assessment approaches are developed such as PID methods. The existing KAERI FEP list was reviewed. Based on these new reference and alternative scenarios are developed along with a new code based on the Goldsim. The code based on the compartment theory can be applied to assess both normal and what if scenarios. In addition detailed studies on THRC coupling is studied. The oriental biosphere study ends with great success over the completion of code V and V with JAEA. The further development of quality assurance, in the form of the CYPRUS+ enables handy use of it for information management

  10. A dynamic Bayesian network based approach to safety decision support in tunnel construction

    International Nuclear Information System (INIS)

    Wu, Xianguo; Liu, Huitao; Zhang, Limao; Skibniewski, Miroslaw J.; Deng, Qianli; Teng, Jiaying

    2015-01-01

    This paper presents a systemic decision approach with step-by-step procedures based on dynamic Bayesian network (DBN), aiming to provide guidelines for dynamic safety analysis of the tunnel-induced road surface damage over time. The proposed DBN-based approach can accurately illustrate the dynamic and updated feature of geological, design and mechanical variables as the construction progress evolves, in order to overcome deficiencies of traditional fault analysis methods. Adopting the predictive, sensitivity and diagnostic analysis techniques in the DBN inference, this approach is able to perform feed-forward, concurrent and back-forward control respectively on a quantitative basis, and provide real-time support before and after an accident. A case study in relating to dynamic safety analysis in the construction of Wuhan Yangtze Metro Tunnel in China is used to verify the feasibility of the proposed approach, as well as its application potential. The relationships between the DBN-based and BN-based approaches are further discussed according to analysis results. The proposed approach can be used as a decision tool to provide support for safety analysis in tunnel construction, and thus increase the likelihood of a successful project in a dynamic project environment. - Highlights: • A dynamic Bayesian network (DBN) based approach for safety decision support is developed. • This approach is able to perform feed-forward, concurrent and back-forward analysis and control. • A case concerning dynamic safety analysis in Wuhan Yangtze Metro Tunnel in China is presented. • DBN-based approach can perform a higher accuracy than traditional static BN-based approach

  11. Risk-informed decision making during Bohunice NPP safety upgrading

    International Nuclear Information System (INIS)

    Lipar, M.; Muzikova, E.; Kubanyi, J.

    2001-01-01

    The paper summarizes some facts of risk-informed regulation developments within UJD regulatory environment. Based on national as well as international operating experience and indications resulted from PSA, Nuclear Regulatory Authority of the Slovak Republic (UJD) since its constituting in 1993 has devoted an effort to use PSA technology to support the regulatory policy in Slovakia. The PSA is considered a complement, not a substitute, to the deterministic approach. Suchlike integrated approach is used in decision making processes and the final decision on scope and priorities is based on it. The paper outlines risk insights used in the decision making process concerning Bohunice NPP safety upgrading and focuses on the role of PSA results in Gradual Reconstruction of Bohunice VI NPP. Besides, two other examples of the PSA results application to the decision making process are provided: the assessment of proposal of modifications to the main power supply diagram (incorporation of generator switches) and the assessment of licensee request for motor generator AOT (Allowable Outage Time) extension. As an example of improving support of Bohunice V-2 risk-informed operations, concept of AOT calculations and Bohunice V-2 Risk Monitor Project are briefly described. (author)

  12. Decision strategy research: system analysis

    International Nuclear Information System (INIS)

    Carle, B.

    2000-01-01

    The objective of SCK-CEN's R and D programme on decision strategies is (1) to develop theories, methods and software tools which help decision makers shape, analyse and understand their decisions; (2) to study group processes in decision making; (3) to apply theories, methods and tools in a context related to nuclear emergency preparedness and more generally to support in a context dealing with ionising radiation; (4) to increase SCK-CEN's knowledge on general emergency preparedness and to introduce SCK-CEN staff to computer supported decision techniques. Ongoing R and D has two components: (1) the study of the use of information and knowledge transfer in group decision processes, and more specific studying important factors when computers are used as information source and communication tool; and (2) the study of preference modelling individually and during group decision processes. Principal achievements in 1999 are described

  13. Decision strategy research: system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Carle, B

    2000-07-01

    The objective of SCK-CEN's R and D programme on decision strategies is (1) to develop theories, methods and software tools which help decision makers shape, analyse and understand their decisions; (2) to study group processes in decision making; (3) to apply theories, methods and tools in a context related to nuclear emergency preparedness and more generally to support in a context dealing with ionising radiation; (4) to increase SCK-CEN's knowledge on general emergency preparedness and to introduce SCK-CEN staff to computer supported decision techniques. Ongoing R and D has two components: (1) the study of the use of information and knowledge transfer in group decision processes, and more specific studying important factors when computers are used as information source and communication tool; and (2) the study of preference modelling individually and during group decision processes. Principal achievements in 1999 are described.

  14. Applying decision trial and evaluation laboratory as a decision tool for effective safety management system in aviation transport

    Directory of Open Access Journals (Sweden)

    Ifeanyichukwu Ebubechukwu Onyegiri

    2016-10-01

    Full Text Available In recent years, in the aviation industry, the weak engineering controls and lapses associated with safety management systems (SMSs are responsible for the seemingly unprecedented disasters. A previous study has confirmed the difficulties experienced by safety managers with SMSs and the need to direct research to this area of investigation for more insights and progress in the evaluation and maintenance of SMSs in the aviation industry. The purpose of this work is to examine the application of Decision Trial and Evaluation Laboratory (DEMATEL to the aviation industry in developing countries with illustration using the Nigerian aviation survey data for the validation of the method. The advantage of the procedure over other decision making methods is in its ability to apply feedback in its decision making. It also affords us the opportunity of breaking down the complex aviation SMS components and elements which are multi-variate in nature through the analysis of the contributions of the diverse system criteria from the perspective of cause and effects, which in turn yields easier and yet more effective aviation transportation accident pre-corrective actions. In this work, six revised components of an SMS were identified and DEMATEL was applied to obtain their direct and indirect impacts and influences on the overall SMS performance. Data collection was by the survey questionnaire, which served as the initial direct-relation matrix, coded in Matlab software for establishing the impact relation map (IRM. The IRM was then plotted in MS Excel spread-sheet software. From our results, safety structure and regulation has the highest impact level on an SMS with a corresponding positive relation level value. In conclusion, the results agree with those of previous researchers that used grey relational analysis. Thus, DEMATEL serves as a great tool and resource for the safety manager.

  15. Binary Decision Tree Development for Probabilistic Safety Assessment Applications

    International Nuclear Information System (INIS)

    Simic, Z.; Banov, R.; Mikulicic, V.

    2008-01-01

    The aim of this article is to describe state of the development for the relatively new approach in the probabilistic safety analysis (PSA). This approach is based on the application of binary decision diagrams (BDD) representation for the logical function on the quantitative and qualitative analysis of complex systems that are presented by fault trees and event trees in the PSA applied for the nuclear power plants risk determination. Even BDD approach offers full solution comparing to the partial one from the conventional quantification approach there are still problems to be solved before new approach could be fully implemented. Major problem with full application of BDD is difficulty of getting any solution for the PSA models of certain complexity. This paper is comparing two approaches in PSA quantification. Major focus of the paper is description of in-house developed BDD application with implementation of the original algorithms. Resulting number of nodes required to represent the BDD is extremely sensitive to the chosen order of variables (i.e., basic events in PSA). The problem of finding an optimal order of variables that form the BDD falls under the class of NP-complete complexity. This paper presents an original approach to the problem of finding the initial order of variables utilized for the BDD construction by various dynamical reordering schemes. Main advantage of this approach compared to the known methods of finding the initial order is with better results in respect to the required working memory and time needed to finish the BDD construction. Developed method is compared against results from well known methods such as depth-first, breadth-first search procedures. Described method may be applied in finding of an initial order for fault trees/event trees being created from basic events by means of logical operations (e.g. negation, and, or, exclusive or). With some testing models a significant reduction of used memory has been achieved, sometimes

  16. Safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Selvatici, E.

    1981-01-01

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

  17. Licensing decisions and safety research related to LMFBR accidents

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. Problems of making decisions with account of risk and safety factors

    Energy Technology Data Exchange (ETDEWEB)

    Larichev, O I

    1987-01-01

    New trends in making decisions on accidents when using large-scale technologies-NPPs, chemical plants etc., are considered. Three main directions in the investigations in this field are distinguished. One of them consists in risk measuring (its perception by people, ways of its quantitative determination). The second direction consists in increasing the safety of large-scale production systems. Here the following questions are considered: risk assessment (the safety standard statement), site selection for new systems, man-machine interaction problems, development of safer technologies, cost benefit safety analysis. The third direction is connected with the problem of accidents and their analysis. This direction includes considering the reasons and process of the accident development, preparing for the possible accidents, monitoring under extreme conditions, accident effect analysis.

  19. Problems of making decisions with account of risk and safety factors

    International Nuclear Information System (INIS)

    Larichev, O.I.

    1987-01-01

    New trends in making decisions on accidents when using large-scale technologies-NPPs, chemical plants etc., are considered. Three main directions in the investigations in this field are distinguished. One of them consists in risk measuring (its perception by people, ways of its quantitative determination). The second direction consists in increasing the safety of large-scale production systems. Here the following questions are considered: risk assessment (the safety standard statement), site selection for new systems, man-machine interaction problems, development of safer technologies, cost benefit safety analysis. The third direction is connected with the problem of accidents and their analysis. This direction includes considering the reasons and process of the accident development, preparing for the possible accidents, monitoring under extreme conditions, accident effect analysis

  20. An analysis of medical decision making

    International Nuclear Information System (INIS)

    Lusted, L.B.

    1977-01-01

    Medical decision-making studies continue to focus on two questions: How do physicians make decisions and how should physicians make decisions. Researchers pursuing the first question emphasize human cognitive processes and the programming of symbol systems to model the observed human behaviour. Those researchers concentrating on the second question assume that there is a standard of performance against which physicians' decisions can be judged, and to help the physician improve his performance an array of tools is proposed. These tools include decision trees, Bayesian analysis, decision matrices, receiver operating characteristic (ROC) analysis, and cost-benefit considerations including utility measures. Both questions must be answered in an ethical context where ethics and decision analysis are intertwined. (author)

  1. The LaSalle probabilistic safety analysis

    International Nuclear Information System (INIS)

    Frederick, L.G.; Massin, H.L.; Crane, G.R.

    1987-01-01

    A probabilistic safety analysis has been performed for LaSalle County Station, a twin-unit General Electric BWR5 Mark II nuclear power plant. A primary objective of this PSA is to provide engineers with a useful and useable tool for making design decisions, performing technical specification optimization, evaluating proposed regulatory changes to equipment and procedures, and as an aid in operator training. Other objectives are to identify the hypothetical accident sequences that would contribute to core damage frequency, and to provide assurance that the total expected frequency of core-damaging accidents is below 10 -4 per reactor-year in response to suggested goals. (orig./HSCH)

  2. Integrating technical analysis and public values in risk-based decision making

    International Nuclear Information System (INIS)

    Bohnenblust, Hans; Slovic, Paul

    1998-01-01

    Simple technical analysis cannot capture the complex scope of preferences or values of society and individuals. However, decision making needs to be sustained by formal analysis. The paper describes a policy framework which incorporates both technical analysis and aspects of public values. The framework can be used as a decision supporting tool and helps decision makers to make more informed and more transparent decisions about safety issues

  3. THE CAUSAL ANALYSIS / DIAGNOSIS DECISION ...

    Science.gov (United States)

    CADDIS is an on-line decision support system that helps investigators in the regions, states and tribes find, access, organize, use and share information to produce causal evaluations in aquatic systems. It is based on the US EPA's Stressor Identification process which is a formal method for identifying causes of impairments in aquatic systems. CADDIS 2007 increases access to relevant information useful for causal analysis and provides methods and tools that practitioners can use to analyze their own data. The new Candidate Cause section provides overviews of commonly encountered causes of impairments to aquatic systems: metals, sediments, nutrients, flow alteration, temperature, ionic strength, and low dissolved oxygen. CADDIS includes new Conceptual Models that illustrate the relationships from sources to stressors to biological effects. An Interactive Conceptual Model for phosphorus links the diagram with supporting literature citations. The new Analyzing Data section helps practitioners analyze their data sets and interpret and use those results as evidence within the USEPA causal assessment process. Downloadable tools include a graphical user interface statistical package (CADStat), and programs for use with the freeware R statistical package, and a Microsoft Excel template. These tools can be used to quantify associations between causes and biological impairments using innovative methods such as species-sensitivity distributions, biological inferenc

  4. Computer aided safety analysis 1989

    International Nuclear Information System (INIS)

    1990-04-01

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

  5. Operating plant safety analysis needs

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  6. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    E.N. Lindner

    2004-12-03

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

  7. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    International Nuclear Information System (INIS)

    E.N. Lindner

    2004-01-01

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

  8. Data Decision Analysis: Project Shoal

    Energy Technology Data Exchange (ETDEWEB)

    Forsgren, Frank; Pohll, Greg; Tracy, John

    1999-01-01

    The purpose of this study was to determine the most appropriate field activities in terms of reducing the uncertainty in the groundwater flow and transport model at the Project Shoal area. The data decision analysis relied on well-known tools of statistics and uncertainty analysis. This procedure identified nine parameters that were deemed uncertain. These included effective porosity, hydraulic head, surface recharge, hydraulic conductivity, fracture correlation scale, fracture orientation, dip angle, dissolution rate of radionuclides from the puddle glass, and the retardation coefficient, which describes the sorption characteristics. The parameter uncertainty was described by assigning prior distributions for each of these parameters. Next, the various field activities were identified that would provide additional information on these parameters. Each of the field activities was evaluated by an expert panel to estimate posterior distribution of the parameters assuming a field activity was performed. The posterior distributions describe the ability of the field activity to estimate the true value of the nine parameters. Monte Carlo techniques were used to determine the current uncertainty, the reduction of uncertainty if a single parameter was known with certainty, and the reduction of uncertainty expected from each field activity on the model predictions. The mean breakthrough time to the downgradient land withdrawal boundary and the peak concentration at the control boundary were used to evaluate the uncertainty reduction. The radionuclide 137Cs was used as the reference solute, as its migration is dependent on all of the parameters. The results indicate that the current uncertainty of the model yields a 95 percent confidence interval between 42 and 1,412 years for the mean breakthrough time and an 18 order-of-magnitude range in peak concentration. The uncertainty in effective porosity and recharge dominates the uncertainty in the model predictions, while the

  9. Decision analysis for the selection of tank waste retrieval technology

    International Nuclear Information System (INIS)

    DAVIS, FREDDIE J.; DEWEESE, GREGORY C.; PICKETT, WILLIAM W.

    2000-01-01

    The objective of this report is to supplement the C-104 Alternatives Generation and Analysis (AGA) by providing a decision analysis for the alternative technologies described therein. The decision analysis used the Multi-Attribute Utility Analysis (MUA) technique. To the extent possible information will come from the AGA. Where data are not available, elicitation of expert opinion or engineering judgment is used and reviewed by the authors of the AGA. A key element of this particular analysis is the consideration of varying perspectives of parties interested in or affected by the decision. The six alternatives discussed are: sluicing; sluicing with vehicle mounted transfer pump; borehole mining; vehicle with attached sluicing nozzle and pump; articulated arm with attached sluicing nozzle; and mechanical dry retrieval. These are evaluated using four attributes, namely: schedule, cost, environmental impact, and safety

  10. Probabilistic Analysis in Management Decision Making

    DEFF Research Database (Denmark)

    Delmar, M. V.; Sørensen, John Dalsgaard

    1992-01-01

    The target group in this paper is people concerned with mathematical economic decision theory. It is shown how the numerically effective First Order Reliability Methods (FORM) can be used in rational management decision making, where some parameters in the applied decision basis are uncertainty...... quantities. The uncertainties are taken into account consistently and the decision analysis is based on the general decision theory in combination with reliability and optimization theory. Examples are shown where the described technique is used and some general conclusion are stated....

  11. Status of safety analysis reports

    International Nuclear Information System (INIS)

    Cserhati, A.

    1999-01-01

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

  12. Status of safety analysis reports

    Energy Technology Data Exchange (ETDEWEB)

    Cserhati, A

    1999-06-01

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

  13. Statistical considerations on safety analysis

    International Nuclear Information System (INIS)

    Pal, L.; Makai, M.

    2004-01-01

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

  14. The Markov Latent Effects Approach to Safety and Decision -Making; TOPICAL

    International Nuclear Information System (INIS)

    COOPER, J. ARLIN

    2001-01-01

    The methodology in this report addresses the safety effects of organizational and operational factors that can be measured through ''inspection.'' The investigation grew out of a preponderance of evidence that the safety ''culture'' (attitude of employees and management toward safety) was frequently one of the major root causes behind accidents or safety-relevant failures. The approach is called ''Markov latent effects'' analysis. Since safety also depends on a multitude of factors that are best measured through well known risk analysis methods (e.g., fault trees, event trees, FMECA, physical response modeling, etc.), the Markov latent effects approach supplements conventional safety assessment and decision analysis methods. A top-down mathematical approach is developed for decomposing systems, for determining the most appropriate items to be measured, and for expressing the measurements as imprecise subjective metrics through possibilistic or fuzzy numbers. A mathematical model is developed that facilitates combining (aggregating) inputs into overall metrics and decision aids, also portraying the inherent uncertainty. A major goal of the modeling is to help convey the top-down system perspective. Metrics are weighted according to significance of the attribute with respect to subsystems and are aggregated nonlinearly. Since the accumulating effect responds less and less to additional contribution, it is termed ''soft'' mathematical aggregation, which is analogous to how humans frequently make decisions. Dependence among the contributing factors is accounted for by incorporating subjective metrics on commonality and by reducing the overall contribution of these combinations to the overall aggregation. Decisions derived from the results are facilitated in several ways. First, information is provided on input ''Importance'' and ''Sensitivity'' (both Primary and Secondary) in order to know where to place emphasis on investigation of root causes and in considering new

  15. Health economics and outcomes methods in risk-based decision-making for blood safety.

    Science.gov (United States)

    Custer, Brian; Janssen, Mart P

    2015-08-01

    Analytical methods appropriate for health economic assessments of transfusion safety interventions have not previously been described in ways that facilitate their use. Within the context of risk-based decision-making (RBDM), health economics can be important for optimizing decisions among competing interventions. The objective of this review is to address key considerations and limitations of current methods as they apply to blood safety. Because a voluntary blood supply is an example of a public good, analyses should be conducted from the societal perspective when possible. Two primary study designs are recommended for most blood safety intervention assessments: budget impact analysis (BIA), which measures the cost to implement an intervention both to the blood operator but also in a broader context, and cost-utility analysis (CUA), which measures the ratio between costs and health gain achieved, in terms of reduced morbidity and mortality, by use of an intervention. These analyses often have important limitations because data that reflect specific aspects, for example, blood recipient population characteristics or complication rates, are not available. Sensitivity analyses play an important role. The impact of various uncertain factors can be studied conjointly in probabilistic sensitivity analyses. The use of BIA and CUA together provides a comprehensive assessment of the costs and benefits from implementing (or not) specific interventions. RBDM is multifaceted and impacts a broad spectrum of stakeholders. Gathering and analyzing health economic evidence as part of the RBDM process enhances the quality, completeness, and transparency of decision-making. © 2015 AABB.

  16. Advancing Alternative Analysis: Integration of Decision Science

    DEFF Research Database (Denmark)

    Malloy, Timothy F; Zaunbrecher, Virginia M; Batteate, Christina

    2016-01-01

    Decision analysis-a systematic approach to solving complex problems-offers tools and frameworks to support decision making that are increasingly being applied to environmental challenges. Alternatives analysis is a method used in regulation and product design to identify, compare, and evaluate......, and civil society and included experts in toxicology, decision science, alternatives assessment, engineering, and law and policy. Participants were divided into two groups and prompted with targeted questions. Throughout the workshop, the groups periodically came together in plenary sessions to reflect......) engaging the systematic development and evaluation of decision approaches and tools; (2) using case studies to advance the integration of decision analysis into alternatives analysis; (3) supporting transdisciplinary research; and (4) supporting education and outreach efforts....

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

    Science.gov (United States)

    Wu, Jun; Li, Chengbing; Huo, Yueying

    2014-01-01

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

  18. Removing unreasonable conservatisms in DOE safety analysis

    International Nuclear Information System (INIS)

    BISHOP, G.E.

    1999-01-01

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

  19. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

    Havlůj, F.; Hejzlar, J.; Vočka, R.

    2013-01-01

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

  20. Decision analysis for deteriorating structures

    International Nuclear Information System (INIS)

    Val, Dimitri V.; Stewart, Mark G.

    2005-01-01

    Measures that improve durability of a structure usually increase its initial cost. Thus, in order to make a decision about a cost-effective solution the life-cycle cost of a structure including cost of structural failure needs to be considered. Due to uncertainties associated with structural properties, loads and environmental conditions the cost of structural failure is a random variable. The paper derives probability distributions of the cost of failure of a single structure and a group of identical structures when single or multiple failures are possible during the service life of a structure. The probability distributions are based on cumulative probabilities of failure of a single structure over its service life. It is assumed that failures occur at discrete points in time, the cost of failure set at the time of decision making remains constant for a particular design solution and the discount rate is a deterministic parameter not changing with time. The probability distributions can be employed to evaluate the expected life-cycle cost or the expected utility, which is then used in decision making. An example, which considers the selection of durability specifications for a reinforced concrete structure built on the coast, illustrates the use of the derived probability distributions

  1. Comparative Analysis of Investment Decision Models

    Directory of Open Access Journals (Sweden)

    Ieva Kekytė

    2017-06-01

    Full Text Available Rapid development of financial markets resulted new challenges for both investors and investment issues. This increased demand for innovative, modern investment and portfolio management decisions adequate for market conditions. Financial market receives special attention, creating new models, includes financial risk management and investment decision support systems.Researchers recognize the need to deal with financial problems using models consistent with the reality and based on sophisticated quantitative analysis technique. Thus, role mathematical modeling in finance becomes important. This article deals with various investments decision-making models, which include forecasting, optimization, stochatic processes, artificial intelligence, etc., and become useful tools for investment decisions.

  2. Systems engineered health and safety criteria for safety analysis reports

    International Nuclear Information System (INIS)

    Beitel, G.A.; Morcos, N.

    1993-01-01

    The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

  3. Conflicts about nuclear power safety: a decision theoretic approach

    International Nuclear Information System (INIS)

    Winterfeldt, D.V.; Rios, M.

    1980-01-01

    A series of psychological studies indicate that people's judgements of risks from energy production in general and nuclear power plants in particular deviate from technical and statistical estimates because social and psychological variables influence people's risk perception. After reviewing these studies a decision analytic methodology is outlined which incorporates such social and psychological variables in a formal analysis of the risks and benefits of nuclear energy production. The methodology is intended to identify groups with differing risk-benefit perceptions and to elicit and quantify their values and concerns. Such group and value structures are presented for the problem of choosing between a nuclear plant, a coal plant, and a conservation strategy

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

    International Nuclear Information System (INIS)

    Demin, V.

    2004-01-01

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

  5. 10 CFR 70.62 - Safety program and integrated safety analysis.

    Science.gov (United States)

    2010-01-01

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

  6. Beyond Decision Making for Outdoor Leaders: Expanding the Safety Behavior Research Agenda

    Science.gov (United States)

    Jackson, Jeff S.

    2016-01-01

    The study of safety behaviour of designated outdoor leaders primarily revolves around their decision making and judgement. The last ten years, however, have seen relatively little peer-reviewed research regarding guide or instructor safety cognition and behaviour. The narrow decision making focus of modern work makes for a field of study…

  7. Periodic safety review of the HTR-10 safety analysis

    International Nuclear Information System (INIS)

    Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

    2015-01-01

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

  8. Transient analysis for resolving safety issues

    International Nuclear Information System (INIS)

    Chao, J.; Layman, W.

    1987-01-01

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

  9. Sensitivity Analysis in Sequential Decision Models.

    Science.gov (United States)

    Chen, Qiushi; Ayer, Turgay; Chhatwal, Jagpreet

    2017-02-01

    Sequential decision problems are frequently encountered in medical decision making, which are commonly solved using Markov decision processes (MDPs). Modeling guidelines recommend conducting sensitivity analyses in decision-analytic models to assess the robustness of the model results against the uncertainty in model parameters. However, standard methods of conducting sensitivity analyses cannot be directly applied to sequential decision problems because this would require evaluating all possible decision sequences, typically in the order of trillions, which is not practically feasible. As a result, most MDP-based modeling studies do not examine confidence in their recommended policies. In this study, we provide an approach to estimate uncertainty and confidence in the results of sequential decision models. First, we provide a probabilistic univariate method to identify the most sensitive parameters in MDPs. Second, we present a probabilistic multivariate approach to estimate the overall confidence in the recommended optimal policy considering joint uncertainty in the model parameters. We provide a graphical representation, which we call a policy acceptability curve, to summarize the confidence in the optimal policy by incorporating stakeholders' willingness to accept the base case policy. For a cost-effectiveness analysis, we provide an approach to construct a cost-effectiveness acceptability frontier, which shows the most cost-effective policy as well as the confidence in that for a given willingness to pay threshold. We demonstrate our approach using a simple MDP case study. We developed a method to conduct sensitivity analysis in sequential decision models, which could increase the credibility of these models among stakeholders.

  10. How decision analysis can further nanoinformatics.

    Science.gov (United States)

    Bates, Matthew E; Larkin, Sabrina; Keisler, Jeffrey M; Linkov, Igor

    2015-01-01

    The increase in nanomaterial research has resulted in increased nanomaterial data. The next challenge is to meaningfully integrate and interpret these data for better and more efficient decisions. Due to the complex nature of nanomaterials, rapid changes in technology, and disunified testing and data publishing strategies, information regarding material properties is often illusive, uncertain, and/or of varying quality, which limits the ability of researchers and regulatory agencies to process and use the data. The vision of nanoinformatics is to address this problem by identifying the information necessary to support specific decisions (a top-down approach) and collecting and visualizing these relevant data (a bottom-up approach). Current nanoinformatics efforts, however, have yet to efficiently focus data acquisition efforts on the research most relevant for bridging specific nanomaterial data gaps. Collecting unnecessary data and visualizing irrelevant information are expensive activities that overwhelm decision makers. We propose that the decision analytic techniques of multicriteria decision analysis (MCDA), value of information (VOI), weight of evidence (WOE), and portfolio decision analysis (PDA) can bridge the gap from current data collection and visualization efforts to present information relevant to specific decision needs. Decision analytic and Bayesian models could be a natural extension of mechanistic and statistical models for nanoinformatics practitioners to master in solving complex nanotechnology challenges.

  11. NKS/SOS-1 seminar on safety analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-01

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

  12. Needs for evidence-based road safety decision making in Europe.

    NARCIS (Netherlands)

    Dupont, E. Muhlrad, N. Buttler, I. Gitelman, V. Giustiniani, G. Jähi, H. Machata, K. Martensen, H. Papadimitriou, E. Persia, L. Talbot, R. Vallet, G. Wijnen, W. & Yannis, G.

    2012-01-01

    The objective of this research is the assessment of current needs for evidence-based road safety decision making in Europe, through the consultation of a panel of road safety experts. The members of this Experts Panel have extensive knowledge of road safety management processes and needs in their

  13. Safety validation of decision trees for hepatocellular carcinoma.

    Science.gov (United States)

    Wang, Xian-Qiang; Liu, Zhe; Lv, Wen-Ping; Luo, Ying; Yang, Guang-Yun; Li, Chong-Hui; Meng, Xiang-Fei; Liu, Yang; Xu, Ke-Sen; Dong, Jia-Hong

    2015-08-21

    To evaluate a different decision tree for safe liver resection and verify its efficiency. A total of 2457 patients underwent hepatic resection between January 2004 and December 2010 at the Chinese PLA General Hospital, and 634 hepatocellular carcinoma (HCC) patients were eligible for the final analyses. Post-hepatectomy liver failure (PHLF) was identified by the association of prothrombin time 50 μmol/L (the "50-50" criteria), which were assessed at day 5 postoperatively or later. The Swiss-Clavien decision tree, Tokyo University-Makuuchi decision tree, and Chinese consensus decision tree were adopted to divide patients into two groups based on those decision trees in sequence, and the PHLF rates were recorded. The overall mortality and PHLF rate were 0.16% and 3.0%. A total of 19 patients experienced PHLF. The numbers of patients to whom the Swiss-Clavien, Tokyo University-Makuuchi, and Chinese consensus decision trees were applied were 581, 573, and 622, and the PHLF rates were 2.75%, 2.62%, and 2.73%, respectively. Significantly more cases satisfied the Chinese consensus decision tree than the Swiss-Clavien decision tree and Tokyo University-Makuuchi decision tree (P decision trees. The Chinese consensus decision tree expands the indications for hepatic resection for HCC patients and does not increase the PHLF rate compared to the Swiss-Clavien and Tokyo University-Makuuchi decision trees. It would be a safe and effective algorithm for hepatectomy in patients with hepatocellular carcinoma.

  14. SIMMER as a safety analysis tool

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  15. Decision Vulnerability Analysis (DVA) Program

    Science.gov (United States)

    2014-05-01

    31 14 Graphical Representation of the Summary Judgments of the Effectiveness, Vulnerability, and Understanding of the Subsystems’ as Judged by...posed several challenges. Numerous organizational typologies have been suggested over the years ( Robbins , 1994), and these typologies are often based...structure and functioning from a typology perspective ( Robbins , 1994), excerpts from a task analysis that described how the analysts currently performed

  16. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

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

  17. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

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

  18. Decisions, decisions: analysis of age, cohort, and time of testing on framing of risky decision options.

    Science.gov (United States)

    Mayhorn, Christopher B; Fisk, Arthur D; Whittle, Justin D

    2002-01-01

    Decision making in uncertain environments is a daily challenge faced by adults of all ages. Framing decision options as either gains or losses is a common method of altering decision-making behavior. In the experiment reported here, benchmark decision-making data collected in the 1970s by Tversky and Kahneman (1981, 1988) were compared with data collected from current samples of young and older adults to determine whether behavior was consistent across time. Although differences did emerge between the benchmark and the present samples, the effect of framing on decision behavior was relatively stable. The present findings suggest that adults of all ages are susceptible to framing effects. Results also indicated that apparent age differences might be better explained by an analysis of cohort and time-of-testing effects. Actual or potential applications of this research include an understanding of how framing might influence the decision-making behavior of people of all ages in a number of applied contexts, such as product warning interactions and medical decision scenarios.

  19. Corporate financial decision-makers' perceptions of workplace safety.

    Science.gov (United States)

    Huang, Yueng-Hsiang; Leamon, Tom B; Courtney, Theodore K; Chen, Peter Y; DeArmond, Sarah

    2007-07-01

    This study, through a random national survey, explored how senior financial executives or managers (those who determined high-level budget, resource allocation, and corporate priorities) of medium-to-large companies perceive important workplace safety issues. The three top-rated safety priorities in resource allocation reported by the participants (overexertion, repetitive motion, and bodily reaction) were consistent with the top three perceived causes of workers' compensation losses. The greatest single safety concerns reported were overexertion, repetitive motion, highway accidents, falling on the same level and bodily reaction. A majority of participants believed that the indirect costs associated with workplace injury were higher than the direct costs. Our participants believed that money spent improving workplace safety would have significant returns. The perceived top benefits of an effective workplace safety program were increased productivity, reduced cost, retention, and increased satisfaction among employees. The perceived most important safety modification was safety training. The top reasons senior financial executives gave for believing their safety programs were better than those at other companies were that their companies paid more attention to and emphasized safety, they had better classes and training focused on safety, and they had teams/individuals focused specifically on safety.

  20. Decision analysis with cumulative prospect theory.

    Science.gov (United States)

    Bayoumi, A M; Redelmeier, D A

    2000-01-01

    Individuals sometimes express preferences that do not follow expected utility theory. Cumulative prospect theory adjusts for some phenomena by using decision weights rather than probabilities when analyzing a decision tree. The authors examined how probability transformations from cumulative prospect theory might alter a decision analysis of a prophylactic therapy in AIDS, eliciting utilities from patients with HIV infection (n = 75) and calculating expected outcomes using an established Markov model. They next focused on transformations of three sets of probabilities: 1) the probabilities used in calculating standard-gamble utility scores; 2) the probabilities of being in discrete Markov states; 3) the probabilities of transitioning between Markov states. The same prophylaxis strategy yielded the highest quality-adjusted survival under all transformations. For the average patient, prophylaxis appeared relatively less advantageous when standard-gamble utilities were transformed. Prophylaxis appeared relatively more advantageous when state probabilities were transformed and relatively less advantageous when transition probabilities were transformed. Transforming standard-gamble and transition probabilities simultaneously decreased the gain from prophylaxis by almost half. Sensitivity analysis indicated that even near-linear probability weighting transformations could substantially alter quality-adjusted survival estimates. The magnitude of benefit estimated in a decision-analytic model can change significantly after using cumulative prospect theory. Incorporating cumulative prospect theory into decision analysis can provide a form of sensitivity analysis and may help describe when people deviate from expected utility theory.

  1. Financial Analysis, Budgeting, Decision and Control

    Directory of Open Access Journals (Sweden)

    Mariana Rodica TIRLEA

    2013-12-01

    Full Text Available The economic processes taking place in the economic environment are stochastic processes that involve and imply risks, arising from product diversification, competition, financial derivatives transactions: swaps, futures, options and from the large number of actors involved in the stock market with a higher or a smaller uncertainty degree. Competition and competitiveness, led to major and rapid change in the business environment, they determined actors participating in the economy to find solutions and methods of collecting and processing data, in such a way that, after being transformed into information they quickly help based on their analysis in decision making, planning and financial forecasting, having an effect on increasing their economic efficiency. In these circumstances the financial analysis, decision, forecasting and control, should be based on quality information that should be a value creation source. The active nature of the financial function implies the existence of a substantially large share of financial analysis, financial decision, forecasting and control.

  2. System for decision analysis support on complex waste management issues

    International Nuclear Information System (INIS)

    Shropshire, D.E.

    1997-01-01

    A software system called the Waste Flow Analysis has been developed and applied to complex environmental management processes for the United States Department of Energy (US DOE). The system can evaluate proposed methods of waste retrieval, treatment, storage, transportation, and disposal. Analysts can evaluate various scenarios to see the impacts to waste slows and schedules, costs, and health and safety risks. Decision analysis capabilities have been integrated into the system to help identify preferred alternatives based on a specific objectives may be to maximize the waste moved to final disposition during a given time period, minimize health risks, minimize costs, or combinations of objectives. The decision analysis capabilities can support evaluation of large and complex problems rapidly, and under conditions of variable uncertainty. The system is being used to evaluate environmental management strategies to safely disposition wastes in the next ten years and reduce the environmental legacy resulting from nuclear material production over the past forty years

  3. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  4. Automation for System Safety Analysis

    Science.gov (United States)

    Malin, Jane T.; Fleming, Land; Throop, David; Thronesbery, Carroll; Flores, Joshua; Bennett, Ted; Wennberg, Paul

    2009-01-01

    This presentation describes work to integrate a set of tools to support early model-based analysis of failures and hazards due to system-software interactions. The tools perform and assist analysts in the following tasks: 1) extract model parts from text for architecture and safety/hazard models; 2) combine the parts with library information to develop the models for visualization and analysis; 3) perform graph analysis and simulation to identify and evaluate possible paths from hazard sources to vulnerable entities and functions, in nominal and anomalous system-software configurations and scenarios; and 4) identify resulting candidate scenarios for software integration testing. There has been significant technical progress in model extraction from Orion program text sources, architecture model derivation (components and connections) and documentation of extraction sources. Models have been derived from Internal Interface Requirements Documents (IIRDs) and FMEA documents. Linguistic text processing is used to extract model parts and relationships, and the Aerospace Ontology also aids automated model development from the extracted information. Visualizations of these models assist analysts in requirements overview and in checking consistency and completeness.

  5. SCOPE safety-controls optimization by performance evaluation: A systematic approach for safety-related decisions at the Hanford Tank Remediation System. Phase 1, final report

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, K.D.; Williams, D.C.; Slezak, S.E.; Young, M.L. [and others

    1996-12-01

    The Department of Energy`s Hanford Tank Waste Remediation system poses a significant challenge for hazard management because of the uncertainty that surrounds many of the variables that must be considered in decisions on safety and control strategies. As a result, site managers must often operate under excessively conservative and expensive assumptions. This report describes a systematic approach to quantifying the uncertainties surrounding the critical parameters in control decisions (e.g., condition of the tanks, kinds of wastes, types of possible accidents) through the use of expert elicitation methods. The results of the elicitations would then be used to build a decision support system and accident analysis model that would allow managers to see how different control strategies would affect the cost and safety of a facility configuration.

  6. SCOPE safety-controls optimization by performance evaluation: A systematic approach for safety-related decisions at the Hanford Tank Remediation System. Phase 1, final report

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Slezak, S.E.; Young, M.L.

    1996-12-01

    The Department of Energy's Hanford Tank Waste Remediation system poses a significant challenge for hazard management because of the uncertainty that surrounds many of the variables that must be considered in decisions on safety and control strategies. As a result, site managers must often operate under excessively conservative and expensive assumptions. This report describes a systematic approach to quantifying the uncertainties surrounding the critical parameters in control decisions (e.g., condition of the tanks, kinds of wastes, types of possible accidents) through the use of expert elicitation methods. The results of the elicitations would then be used to build a decision support system and accident analysis model that would allow managers to see how different control strategies would affect the cost and safety of a facility configuration

  7. Safety analysis - current and future regulatory challenges

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  8. Safety analysis - current and future regulatory challenges

    International Nuclear Information System (INIS)

    Jamieson, T.

    2015-01-01

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

  9. Manpower analysis in transportation safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-05-01

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

  10. Strategic decision analysis applied to borehole seismology

    International Nuclear Information System (INIS)

    Menke, M.M.; Paulsson, B.N.P.

    1994-01-01

    Strategic Decision Analysis (SDA) is the evolving body of knowledge on how to achieve high quality in the decision that shapes an organization's future. SDA comprises philosophy, process concepts, methodology, and tools for making good decisions. It specifically incorporates many concepts and tools from economic evaluation and risk analysis. Chevron Petroleum Technology Company (CPTC) has applied SDA to evaluate and prioritize a number of its most important and most uncertain R and D projects, including borehole seismology. Before SDA, there were significant issues and concerns about the value to CPTC of continuing to work on borehole seismology. The SDA process created a cross-functional team of experts to structure and evaluate this project. A credible economic model was developed, discrete risks and continuous uncertainties were assessed, and an extensive sensitivity analysis was performed. The results, even applied to a very restricted drilling program for a few years, were good enough to demonstrate the value of continuing the project. This paper explains the SDA philosophy concepts, and process and demonstrates the methodology and tools using the borehole seismology project example. SDA is useful in the upstream industry not just in the R and D/technology decisions, but also in major exploration and production decisions. Since a major challenge for upstream companies today is to create and realize value, the SDA approach should have a very broad applicability

  11. 14 CFR 33.75 - Safety analysis.

    Science.gov (United States)

    2010-01-01

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

  12. 14 CFR 35.15 - Safety analysis.

    Science.gov (United States)

    2010-01-01

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

  13. Medical decision making tools: Bayesian analysis and ROC analysis

    International Nuclear Information System (INIS)

    Lee, Byung Do

    2006-01-01

    During the diagnostic process of the various oral and maxillofacial lesions, we should consider the following: 'When should we order diagnostic tests? What tests should be ordered? How should we interpret the results clinically? And how should we use this frequently imperfect information to make optimal medical decision?' For the clinicians to make proper judgement, several decision making tools are suggested. This article discusses the concept of the diagnostic accuracy (sensitivity and specificity values) with several decision making tools such as decision matrix, ROC analysis and Bayesian analysis. The article also explain the introductory concept of ORAD program

  14. AST-500 safety analysis experience

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  15. Management of safety and quality and the relationship with employee decisions in country grain elevators.

    Science.gov (United States)

    Mosher, G A; Keren, N; Freeman, S A; Hurburgh, C R

    2012-07-01

    Human factors play an important role in the management of safety and quality in an agricultural work environment. Although employee actions and decisions have been identified as a key component of successful occupational safety programs and quality management programs, little attention has been given to the employees' role in these types of programs. This research explored two safety relationships that have theoretical connections but little previous research: the relationship between safety climate and quality climate, and the relationship of the safety and quality climates between the organizational level and the group level within a workplace. Survey data were collected at three commercial grain handling facilities from 177 employees. Employees also participated in safety and quality decision-making simulations. Significant positive predictions were noted for safety and quality climate. Decision-making predictions are also discussed. This research suggests that organizational safety is an important predictor of group safety. In addition, recognizing the larger role that supervisors play in group workplace behavior, more should be done to increase employee perceptions of group-level involvement in quality climate to promote more quality-oriented decision-making by employees.

  16. Best-estimate analysis and decision making under uncertainty

    International Nuclear Information System (INIS)

    Orechwa, Y.

    2004-01-01

    In many engineering analyses of system safety the traditional reliance on conservative evaluation model calculations is being replaced with so called best-estimate analysis. These best-estimate analyses differentiate themselves from the traditional conservative analyses through two ingredients, namely realistic models and an account of the residual uncertainty associated with the model calculations. Best-estimate analysis, in the context of this paper, refers to the numerical evaluation of system properties of interest in situations where direct confirmatory measurements are not feasible. A decision with regard to the safety of the system is then made based on the computed numerical values of the system properties of interest. These situations generally arise in the design of systems that require computed and generally nontrivial extrapolations from the available data. In the case of nuclear reactors, examples are criticality of spent fuel pools, neutronic parameters of new advanced designs where insufficient material is available for mockup critical experiments and, the large break loss of coolant accident (LOCA). In this paper the case of LOCA, is taken to discuss the best-estimate analysis and decision making. Central to decision making is information. Thus, of interest is the source, quantity and quality of the information obtained in a best-estimate analysis, and used to define the acceptance criteria and to formulate a decision rule. This in effect expands the problem from the calculation of a conservative margin to a predefined acceptance criterion, to the formulation of a consistent decision rule and the computation of a test statistic for application of the decision rule. The latter view is a necessary condition for developing risk informed decision rules, and, thus, the relation between design basis analysis criteria and probabilistic risk assessment criteria is key. The discussion is in the context of making a decision under uncertainty for a reactor

  17. LMFBR safety experiment facility planning and analysis

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Scott, J.H.

    1976-01-01

    In the past two years considerable effort has been placed on the planning and design of new facilities for the resolution of LMFBR safety issues. The paper reviews the key issues, the experiments needed to resolve them, and the design aspects of proposed new facilities. In addition, it presents a decision theory approach to selecting an optimal combination of modified and new facilities

  18. Nurses' perspectives on the intersection of safety and informed decision making in maternity care.

    Science.gov (United States)

    Jacobson, Carrie H; Zlatnik, Marya G; Kennedy, Holly Powell; Lyndon, Audrey

    2013-01-01

    To explore maternity nurses' perceptions of women's informed decision making during labor and birth to better understand how interdisciplinary communication challenges might affect patient safety. Constructivist grounded theory. Four hospitals in the western United States. Forty-six (46) nurses and physicians practicing in maternity units. Data collection strategies included individual interviews and participant observation. Data were analyzed using the constant comparative method, dimensional analysis, and situational analysis (Charmaz, 2006; Clarke, 2005; Schatzman, 1991). The nurses' central action of holding off harm encompassed three communication strategies: persuading agreement, managing information, and coaching of mothers and physicians. These strategies were executed in a complex, hierarchical context characterized by varied practice patterns and relationships. Nurses' priorities and patient safety goals were sometimes misaligned with those of physicians, resulting in potentially unsafe communication. The communication strategies nurses employed resulted in intended and unintended consequences with safety implications for mothers and providers and had the potential to trap women in the middle of interprofessional conflicts and differences of opinion. © 2013 AWHONN, the Association of Women's Health, Obstetric and Neonatal Nurses.

  19. an analysis of perceived prominent decision making areas in ...

    African Journals Online (AJOL)

    p2333147

    Keywords: Game ranch management, decision making, risk perception, springbuck. ABSTRACT ..... environment, herd management (herd structure) and marketing and client satisfaction .... Prospect theory: An analysis of decision under risk.

  20. Simulation modeling and analysis in safety. II

    International Nuclear Information System (INIS)

    Ayoub, M.A.

    1981-01-01

    The paper introduces and illustrates simulation modeling as a viable approach for dealing with complex issues and decisions in safety and health. The author details two studies: evaluation of employee exposure to airborne radioactive materials and effectiveness of the safety organization. The first study seeks to define a policy to manage a facility used in testing employees for radiation contamination. An acceptable policy is one that would permit the testing of all employees as defined under regulatory requirements, while not exceeding available resources. The second study evaluates the relationship between safety performance and the characteristics of the organization, its management, its policy, and communication patterns among various functions and levels. Both studies use models where decisions are reached based on the prevailing conditions and occurrence of key events within the simulation environment. Finally, several problem areas suitable for simulation studies are highlighted. (Auth.)

  1. Preliminary Integrated Safety Analysis Status Report

    International Nuclear Information System (INIS)

    Gwyn, D.

    2001-01-01

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

  2. Health economics and outcomes methods in risk-based decision-making for blood safety

    NARCIS (Netherlands)

    Custer, Brian; Janssen, Mart P.

    2015-01-01

    Analytical methods appropriate for health economic assessments of transfusion safety interventions have not previously been described in ways that facilitate their use. Within the context of risk-based decision-making (RBDM), health economics can be important for optimizing decisions among competing

  3. Hot Cell Facility (HCF) Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  4. Hot Cell Facility (HCF) Safety Analysis Report

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  5. Approach to uncertainty evaluation for safety analysis

    International Nuclear Information System (INIS)

    Ogura, Katsunori

    2005-01-01

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

  6. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

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

    2000-05-01

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

  7. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  8. Towards a decision support system for control of multiple food safety hazards in raw milk production

    NARCIS (Netherlands)

    Spiegel, van der M.; Sterrenburg, P.; Haasnoot, W.; Fels-Klerx, van der H.J.

    2013-01-01

    Decision support systems (DSS) for controlling multiple food safety hazards in raw milk production have not yet been developed, but the underlying components are fragmentarily available. This article presents the state-of-the-art of essential DSS elements for judging food safety compliance of raw

  9. Mapping a Research Agenda for Home Care Safety: Perspectives from Researchers, Providers, and Decision Makers

    Science.gov (United States)

    Macdonald, Marilyn; Lang, Ariella; MacDonald, Jo-Anne

    2011-01-01

    The purpose of this qualitative interpretive design was to explore the perspectives of researchers, health care providers, policy makers, and decision makers on key risks, concerns, and emerging issues related to home care safety that would inform a line of research inquiry. Defining safety specifically in this home care context has yet to be…

  10. Impediments for the application of risk-informed decision making in nuclear safety

    International Nuclear Information System (INIS)

    Hahn, L.

    2001-01-01

    A broad application of risk-informed decision making in the regulation of safety of nuclear power plants is hindered by the lack of quantitative risk and safety standards as well as of precise instruments to demonstrate an appropriate safety. An additional severe problem is associated with the difficulty to harmonize deterministic design requirements and probabilistic safety assessment. The problem is strengthened by the vulnerability of PSA for subjective influences and the potential of misuse. Beside this scepticism the nuclear community is encouraged to intensify the efforts to improve the quality standards for probabilistic safety assessments and their quality assurance. A prerequisite for reliable risk-informed decision making processes is also a well-defined and transparent relationship between deterministic and probabilistic safety approaches. (author)

  11. Reactor safety impact of functional test intervals: an application of Bayesian decision theory

    International Nuclear Information System (INIS)

    Buoni, F.B.

    1978-01-01

    Functional test intervals for important nuclear reactor systems can be obtained by viewing safety assessment as a decision process and functional testing as a Bayesian learning or information process. A preposterior analysis is used as the analytical model to find the preposterior expected reliability of a system as a function of test intervals. Persistent and transitory failure models are shown to yield different results. Functional tests of systems subject to persistent failure are effective in maintaining system reliability goals. Functional testing is not effective for systems subject to transitory failure; preventive maintenance must be used. A Bayesian posterior analysis of testing data can discriminate between persistent and transitory failure. The role of functional testing is seen to be an aid in assessing the future performance of reactor systems

  12. A regret theory approach to decision curve analysis: A novel method for eliciting decision makers' preferences and decision-making

    OpenAIRE

    Vickers Andrew; Hozo Iztok; Tsalatsanis Athanasios; Djulbegovic Benjamin

    2010-01-01

    Abstract Background Decision curve analysis (DCA) has been proposed as an alternative method for evaluation of diagnostic tests, prediction models, and molecular markers. However, DCA is based on expected utility theory, which has been routinely violated by decision makers. Decision-making is governed by intuition (system 1), and analytical, deliberative process (system 2), thus, rational decision-making should reflect both formal principles of rationality and intuition about good decisions. ...

  13. Updated safety analysis of ITER

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. Updated safety analysis of ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  15. Perspective on safety case to support a possible site recommendation decision

    International Nuclear Information System (INIS)

    Gil, A.V.; Gamble, R.P.

    2002-01-01

    The mission of the US Department of Energy (DOE) is to provide the basis for a national decision regarding the development of a geological repository for spent nuclear fuel and high-level radioactive waste at the Yucca Mountain site in Nevada. There are a number of steps in the decision process defined by US law that must be completed prior to development of a repository at this site. The DOE's focus is currently on the first two steps in this process: characterization of the site to support a determination by the DOE on whether the site is suitable for a geologic repository and a decision by the Secretary of Energy (the Secretary) on whether to recommend to the President that the site be approved for a repository. To enhance the confidence of multiple audiences in the basis for these actions, and to provide a basis for subsequent action by the President and the US Congress, information supporting the decision process must include the elements of a safety case consistent with the statutory and regulatory framework for these decisions. The idea of a safety case is to broaden the basis for confidence by decision-makers and the public in conclusions about safety. A safety case should cite multiple lines of evidence, or reasoning, beyond the results of a safety assessment to support the demonstration of safety, which includes compliance with applicable safety criteria. The multiple lines of evidence should show the basis for confidence in safety. To be most effective, such evidence requires information not directly used in the safety assessment. (author)

  16. Applications of noise analysis to nuclear safety

    International Nuclear Information System (INIS)

    Aguilar Martinez, Omar

    2000-01-01

    Noise Analysis techniques (analysis of the fluctuation of physical parameters) have been successfully applied to the operational vigilance of the technical equipment that plays a decisive role in the production cycle of a very complex industry. Although fluctuation measurements in nuclear installations started almost at the start of the nuclear era (see works by Feynman and Rossi on the development of neutron methodology), only recently have neutron noise diagnostic applications begun to be a part of the standard procedures for the performance of some modern nuclear installations. Following the relevant technical advances made in information sciences and analogical electronics, measuring the fluctuation of physical parameters has become a very effective tool for detecting, guarding and following up possible defects in a nuclear system. As the processing techniques for the fluctuation of a nuclear reactor's physical-neutron parameters have evolved (temporal and frequency analysis, multi-parameter self -regression analysis, etc.), the applications of the theory of non-lineal dynamics and chaos theory have progressed by focusing on the problem from another perspective. This work reports on those nuclear applications of noise analysis that increase nuclear safety in all types of nuclear facilities and that have been carried out by the author over the last decade, such as: -Void Force Critical Set Applications (Zero Power Reactor Applications, Central Institute of Physical Research, Budapest, Hungary); -Research Reactor Applications (Triga Mark III Reactor, National Institute of Nuclear Research, ININ, Mexico); -Power Reactor Applications in a Nuclear Power Plant (First Circuit of Block II, Paks Nuclear Center, Hungary); -Second Loop applications in a Nuclear Power Plant (Block I Paks Nuclear Center, Hungary; Block II Kalinin Nuclear Center, Russia); -Shield System Applications for the Transport of Radioisotopes (Nuclear Technology Center, Havana, Cuba) New trends in

  17. Infusing Reliability Techniques into Software Safety Analysis

    Science.gov (United States)

    Shi, Ying

    2015-01-01

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

  18. Decisions on the safety of using nuclear power

    International Nuclear Information System (INIS)

    Janka, P.

    1992-01-01

    A new nuclear energy law came into force in Finland in 1988. This law defines general principles, conditions and requirements concerning the use of nuclear power. The law expects the use of nuclear power to be safe and the safety and contingency systems to be sufficient. General rules for the safety of using nuclear power and for safety arrangements and contingency plans are laid down by the government. The Finnish Centre for Radiation and Nuclear Safety has proposed the various rules to be adopted by the government and come into force by 1991. The rules for the safety of nuclear power plants and final waste storage plants contain limits for emissions of radioactive substances and radiation exposure and requirements for the safety in planning, building and using nuclear power plants and final waste storage plants. They observe international experience and research on risks linked to the use of nuclear power from the last few years as well as means and measures to contain these risks under all conditions. Safety arrangements at nuclear power plants contain measures required to be taken by the owner of the plants to thwart unlawful activities aimed at the plant. Most important of these are the rules for actions to be taken in dangerous situations. The proposed contingency plans contain measures to be taken by the owner of the plants in order to contain nuclear damages resulting from an accident. Most important of these are the rules for planning contingency arrangements, keeping these arrangements operable and actions to be taken in emergency situations. (author)

  19. Uncertainty analysis in safety assessment

    International Nuclear Information System (INIS)

    Lemos, Francisco Luiz de; Sullivan, Terry

    1997-01-01

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

  20. Chemical plant innovative safety investments decision-support methodology.

    Science.gov (United States)

    Reniers, G L L; Audenaert, A

    2009-01-01

    This article examines the extent to which investing in safety during the creation of a new chemical installation proves profitable. The authors propose a management supporting cost-benefit model that identifies and evaluates investments in safety within a chemical company. This innovative model differentiates between serious accidents and less serious accidents, thus providing an authentic image of prevention-related costs and benefits. In classic cost-benefit analyses, which do not make such differentiations, only a rudimentary image of potential profitability resulting from investments in safety is obtained. The resulting management conclusions that can be drawn from such classical analyses are of a very limited nature. The proposed model, however, is applied to a real case study and the proposed investments in safety at an appointed chemical installation are weighed against the estimated hypothetical benefits resulting from the preventive measures to be installed at the installation. In the case-study carried out in question, it would appear that the proposed prevention investments are justified. Such an economic exercise may be very important to chemical corporations trying to (further) improve their safety investments.

  1. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-11-01

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

  2. Computerised clinical decision support systems to improve medication safety in long-term care homes: a systematic review.

    Science.gov (United States)

    Marasinghe, Keshini Madara

    2015-05-12

    Computerised clinical decision support systems (CCDSS) are used to improve the quality of care in various healthcare settings. This systematic review evaluated the impact of CCDSS on improving medication safety in long-term care homes (LTC). Medication safety in older populations is an important health concern as inappropriate medication use can elevate the risk of potentially severe outcomes (ie, adverse drug reactions, ADR). With an increasing ageing population, greater use of LTC by the growing ageing population and increasing number of medication-related health issues in LTC, strategies to improve medication safety are essential. Databases searched included MEDLINE, EMBASE, Scopus and Cochrane Library. Three groups of keywords were combined: those relating to LTC, medication safety and CCDSS. One reviewer undertook screening and quality assessment. Overall findings suggest that CCDSS in LTC improved the quality of prescribing decisions (ie, appropriate medication orders), detected ADR, triggered warning messages (ie, related to central nervous system side effects, drug-associated constipation, renal insufficiency) and reduced injury risk among older adults. CCDSS have received little attention in LTC, as attested by the limited published literature. With an increasing ageing population, greater use of LTC by the ageing population and increased workload for health professionals, merely relying on physicians' judgement on medication safety would not be sufficient. CCDSS to improve medication safety and enhance the quality of prescribing decisions are essential. Analysis of review findings indicates that CCDSS are beneficial, effective and have potential to improve medication safety in LTC; however, the use of CCDSS in LTC is scarce. Careful assessment on the impact of CCDSS on medication safety and further modifications to existing CCDSS are recommended for wider acceptance. Due to scant evidence in the current literature, further research on implementation and

  3. Independence in regulatory decision making - INSAG-17. A report by the International Nuclear Safety Advisory Group

    International Nuclear Information System (INIS)

    2003-01-01

    This report is intended to promote a common understanding among legislators and other political decision makers, nuclear safety regulators and licensees of the concept of independence in regulatory decision making and how to achieve it. Other interest groups, such as non-governmental organizations and members of the public interested in the regulation of nuclear safety, may also find the report useful. The principles concerning the independence of regulatory organizations are developed and discussed in publications in the IAEA's Safety Standards Series. Although the principles relating to protecting the independence of the regulatory body provide the necessary basis for independence in regulatory decision making, there are additional factors and features that require attention to ensure independence in the decision making by the regulatory body. This INSAG report highlights and discusses a number of such factors and features

  4. Decision Analysis: Engineering Science or Clinical Art

    Science.gov (United States)

    1979-11-01

    TECHNICAL REPORT TR 79-2-97 DECISION ANALYSIS: ENGINEERING SCIENCE OR CLINICAL ART ? by Dennis M. Buede Prepared for Defense Advanced Research...APPLICATIONS OF THE ENGINEER- ING SCIENCE AND CLINICAL ART EXTREMES 9 3.1 Applications of the Engineering Science Approach 9 3.1.1 Mexican electrical...DISCUSSION 29 4.1 Engineering Science versus Clinical Art : A Characterization of When Each is Most Attractive 30 4.2 The Implications of the Engineering

  5. Defense against nuclear weapons: a decision analysis

    International Nuclear Information System (INIS)

    Orient, J.M.

    1985-01-01

    Response to the public health threat posed by nuclear weapons is a medical imperative. The United States, in contrast to other nations, has chosen a course that assures maximal casualties in the event of a nuclear attack, on the theory that prevention of the attack is incompatible with preventive measures against its consequences, such as blast injuries and radiation sickness. A decision analysis approach clarifies the risks and benefits of a change to a strategy of preparedness

  6. Review of decision methodologies for evaluating regulatory actions affecting public health and safety

    International Nuclear Information System (INIS)

    Hendrickson, P.L.; McDonald, C.L.; Schilling, A.H.

    1976-12-01

    This report examines several aspects of the problems and choices facing the governmental decision maker who must take regulatory actions with multiple decision objectives and attributes. Particular attention is given to the problems facing the U.S. Nuclear Regulatory Commission (NRC) and to the decision attribute of chief concern to NRC, the protection of human health and safety, with emphasis on nuclear power plants. The study was undertaken to provide background information for NRC to use in refining its process of value/impact assessment of proposed regulatory actions. The principal conclusion is that approaches to rationally consider the value and impact of proposed regulatory actions are available. These approaches can potentially improve the decision-making process and enable the agency to better explain and defend its decisions. They also permit consistent examination of the impacts, effects of uncertainty and sensitivity to various assumptions of the alternatives being considered. Finally, these approaches can help to assure that affected parties are heard and that technical information is used appropriately and to the extent possible. The principal aspects of the regulatory decision problem covered in the report are: the legal setting for regulatory decisions which affect human health and safety, elements of the decision-making process, conceptual approaches to decision making, current approaches to decision making in several Federal agencies, and the determination of acceptable risk levels

  7. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

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

  8. Towards an integrated approach in supporting microbiological food safety decisions

    DEFF Research Database (Denmark)

    Havelaar, A.H.; Braunig, J.; Christiansen, K.

    2007-01-01

    an integrated scientific approach combining veterinary and medical epidemiology, risk assessment for the farm-to-fork food chain as well as agricultural and health economy. Scientific advice is relevant in all stages of the policy cycle: to assess the magnitude of the food safety problem, to define...

  9. Autoclave nuclear criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  10. Uncertainty analysis in safety assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  11. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

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

  12. Decision Analysis Tools for Volcano Observatories

    Science.gov (United States)

    Hincks, T. H.; Aspinall, W.; Woo, G.

    2005-12-01

    Staff at volcano observatories are predominantly engaged in scientific activities related to volcano monitoring and instrumentation, data acquisition and analysis. Accordingly, the academic education and professional training of observatory staff tend to focus on these scientific functions. From time to time, however, staff may be called upon to provide decision support to government officials responsible for civil protection. Recognizing that Earth scientists may have limited technical familiarity with formal decision analysis methods, specialist software tools that assist decision support in a crisis should be welcome. A review is given of two software tools that have been under development recently. The first is for probabilistic risk assessment of human and economic loss from volcanic eruptions, and is of practical use in short and medium-term risk-informed planning of exclusion zones, post-disaster response, etc. A multiple branch event-tree architecture for the software, together with a formalism for ascribing probabilities to branches, have been developed within the context of the European Community EXPLORIS project. The second software tool utilizes the principles of the Bayesian Belief Network (BBN) for evidence-based assessment of volcanic state and probabilistic threat evaluation. This is of practical application in short-term volcano hazard forecasting and real-time crisis management, including the difficult challenge of deciding when an eruption is over. An open-source BBN library is the software foundation for this tool, which is capable of combining synoptically different strands of observational data from diverse monitoring sources. A conceptual vision is presented of the practical deployment of these decision analysis tools in a future volcano observatory environment. Summary retrospective analyses are given of previous volcanic crises to illustrate the hazard and risk insights gained from use of these tools.

  13. Safety analysis in subsurface repositories

    International Nuclear Information System (INIS)

    1985-06-01

    The development of mathematical models to represent the repository-geosphere-biosphere system, and the development of a structure for data acquisition, processing, and use to analyse the safety of subsurface repositories, are presented. To study the behavior of radionuclides in geosphere a laboratory to determine the hydrodynamic dispersion coefficient was constructed. (M.C.K.) [pt

  14. HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

    Energy Technology Data Exchange (ETDEWEB)

    EVANS, C B

    2004-12-21

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

  15. Decision analysis for INEL hazardous waste storage

    Energy Technology Data Exchange (ETDEWEB)

    Page, L.A.; Roach, J.A.

    1994-01-01

    In mid-November 1993, the Idaho National Engineering Laboratory (INEL) Waste Reduction Operations Complex (WROC) Manager requested that the INEL Hazardous Waste Type Manager perform a decision analysis to determine whether or not a new Hazardous Waste Storage Facility (HWSF) was needed to store INEL hazardous waste (HW). In response to this request, a team was formed to perform a decision analysis for recommending the best configuration for storage of INEL HW. Personnel who participated in the decision analysis are listed in Appendix B. The results of the analysis indicate that the existing HWSF is not the best configuration for storage of INEL HW. The analysis detailed in Appendix C concludes that the best HW storage configuration would be to modify and use a portion of the Waste Experimental Reduction Facility (WERF) Waste Storage Building (WWSB), PBF-623 (Alternative 3). This facility was constructed in 1991 to serve as a waste staging facility for WERF incineration. The modifications include an extension of the current Room 105 across the south end of the WWSB and installing heating, ventilation, and bay curbing, which would provide approximately 1,600 ft{sup 2} of isolated HW storage area. Negotiations with the State to discuss aisle space requirements along with modifications to WWSB operating procedures are also necessary. The process to begin utilizing the WWSB for HW storage includes planned closure of the HWSF, modification to the WWSB, and relocation of the HW inventory. The cost to modify the WWSB can be funded by a reallocation of funding currently identified to correct HWSF deficiencies.

  16. Decision analysis for INEL hazardous waste storage

    International Nuclear Information System (INIS)

    Page, L.A.; Roach, J.A.

    1994-01-01

    In mid-November 1993, the Idaho National Engineering Laboratory (INEL) Waste Reduction Operations Complex (WROC) Manager requested that the INEL Hazardous Waste Type Manager perform a decision analysis to determine whether or not a new Hazardous Waste Storage Facility (HWSF) was needed to store INEL hazardous waste (HW). In response to this request, a team was formed to perform a decision analysis for recommending the best configuration for storage of INEL HW. Personnel who participated in the decision analysis are listed in Appendix B. The results of the analysis indicate that the existing HWSF is not the best configuration for storage of INEL HW. The analysis detailed in Appendix C concludes that the best HW storage configuration would be to modify and use a portion of the Waste Experimental Reduction Facility (WERF) Waste Storage Building (WWSB), PBF-623 (Alternative 3). This facility was constructed in 1991 to serve as a waste staging facility for WERF incineration. The modifications include an extension of the current Room 105 across the south end of the WWSB and installing heating, ventilation, and bay curbing, which would provide approximately 1,600 ft 2 of isolated HW storage area. Negotiations with the State to discuss aisle space requirements along with modifications to WWSB operating procedures are also necessary. The process to begin utilizing the WWSB for HW storage includes planned closure of the HWSF, modification to the WWSB, and relocation of the HW inventory. The cost to modify the WWSB can be funded by a reallocation of funding currently identified to correct HWSF deficiencies

  17. Life extension decision making of safety critical systems: An overview

    OpenAIRE

    Shafiee, Mahmood; Animah, I.

    2017-01-01

    In recent years, the concept of “asset life extension” has become increasingly important to safety critical industries including nuclear power, offshore oil and gas, petrochemical, renewable energy, rail transport, aviation, shipping, electricity distribution and transmission, etc. Extending the service life of industrial assets can offer a broad range of economic, technical, social and environmental benefits as compared to other end-of-life management strategies such as decommissioning and r...

  18. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  19. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

    2008-12-01

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

  20. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  1. Shared decision-making during surgical consultation for gallstones at a safety-net hospital.

    Science.gov (United States)

    Mueck, Krislynn M; Leal, Isabel M; Wan, Charlie C; Goldberg, Braden F; Saunders, Tamara E; Millas, Stefanos G; Liang, Mike K; Ko, Tien C; Kao, Lillian S

    2018-04-01

    Understanding patient perspectives regarding shared decision-making is crucial to providing informed, patient-centered care. Little is known about perceptions of vulnerable patients regarding shared decision-making during surgical consultation. The purpose of this study was to evaluate whether a validated tool reflects perceptions of shared decision-making accurately among patients seeking surgical consultation for gallstones at a safety-net hospital. A mixed methods study was conducted in a sample of adult patients with gallstones evaluated at a safety-net surgery clinic between May to July 2016. Semi-structured interviews were conducted after their initial surgical consultation and analyzed for emerging themes. Patients were administered the Shared Decision-Making Questionnaire and Autonomy Preference Scale. Univariate analyses were performed to identify factors associated with shared decision-making and to compare the results of the surveys to those of the interviews. The majority of patients (N = 30) were female (90%), Hispanic (80%), Spanish-speaking (70%), and middle-aged (45.7 ± 16 years). The proportion of patients who perceived shared decision-making was greater in the Shared Decision-Making Questionnaire versus the interviews (83% vs 27%, P decision for operation was not associated with shared decision-making. Contributory factors to this discordance include patient unfamiliarity with shared decision-making, deference to surgeon authority, lack of discussion about different treatments, and confusion between aligned versus shared decisions. Available questionnaires may overestimate shared decision-making in vulnerable patients suggesting the need for alternative or modifications to existing methods. Furthermore, such metrics should be assessed for correlation with patient-reported outcomes, such as satisfaction with decisions and health status. Copyright © 2017 Elsevier Inc. All rights reserved.

  2. Software safety analysis practice in installation phase

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  3. Software safety analysis practice in installation phase

    International Nuclear Information System (INIS)

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

    2010-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

    Kim, Yun Goo; Oh, Eung Se

    2016-01-01

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

  6. Airline Safety: A Comparative Analysis.

    Science.gov (United States)

    1987-01-01

    shrinking FAA inspector force handling a rapidly growing number of air carriers. These studies have always shown an improvement in airline safety in the...EhCLhEmohhhhhhhEoo EhhshhEEmhhhhE EhhEohhEshhhhE EhhhEEEohEohEE EohEEEmhshEmhE IEEE...mmmo 1-2. jI. Mi6 NEW - secuRily CLASSIFICATION OF THIS PAGE (When De

  7. Safety analysis and related studies

    International Nuclear Information System (INIS)

    Lelievre, J.

    1979-12-01

    Several examples of reactor safety studies are given. For light water reactors, the consequences of loss of coolant, the disposition of the fuel elements and the behaviour under irradiation of the steels used for containment are described. For fast reactors, the disposition of fuel elements in the case of cooling accidents and sodium fies are described. Examples given of studies not specific to a particular reactor type include studies of non-destructive testing and those of reliability

  8. Using Addenda in Documented Safety Analysis Reports

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  9. MSSV Modeling for Wolsong-1 Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  10. [HEALTH ECONOMIC ANALYSIS AND FAIR DECISION MAKING].

    Science.gov (United States)

    Jeantet, Marine; Lopez, Alain

    2015-09-01

    Health technology assessment consists in evaluating the incremental cost-benefit ratio of a medicine, a medical device, a vaccine, a health strategy, in comparison to alternative health technologies. This form of socio-eoonomic evaluation aims at optimizing resource allocation within the health system. By setting the terms of valid alternatives, it is useful to highlight public choices, but it cannot in itself make the decision as regards the public funding of patient's access to the considered technology. The decision to include such technology in the basket of health goods and sercices covered, the levels and conditions of the coverage, also result from budget constraints, from economic situation and from a political vision about health policy, social protection and public expenditure. Accordingly, health economic analysis must be implemented on specific and targeted topics. The decision making process, with its health, economic and ethical stakes, calls for a public procedure and debate, based on shared information and argument. Otherwise, health system regulation, confronted with radical and costly innovations in the coming years, will become harder to handle. This requires the development of health economic research teams able to contribute to this assessment exercise.

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

    International Nuclear Information System (INIS)

    1989-10-01

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

  12. Safety analysis of autonomous excavator functionality

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  13. Status of Ignalina's safety analysis reports

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Ignalina NPP is unique among RBMK type reactors in the scope and comprehensiveness of international studies which have been performed to verify its design parameters and analyze risk levels. International assistance took several forms, a very valuable mod of assistance utilized the knowledge of international experts in extensive international studies whose purpose was: collection, systematization and verification of plant design data; analysis of risk levels; recommendations leading to improvements in the safety lave; transfer of state of the art analytical methodology to Lithuanian specialists. The major large scale international studies include: probabilistic risk analysis; extensive international study meant to provide comprehensive overview of plant status with special emphasis on safety aspects; an extensive review of the Safety Analysis Report by an independent group of international experts. In spite of the safety improvements and analyses which have been performed at the Ignalina NPP, much remains to be done in the nearest future

  14. Software safety analysis application in installation phase

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Gumley, P.; Inamdar, S.V.

    1985-01-01

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

  16. Probabilistic safety analysis using microcomputer

    International Nuclear Information System (INIS)

    Futuro Filho, F.L.F.; Mendes, J.E.S.; Santos, M.J.P. dos

    1990-01-01

    The main steps of execution of a Probabilistic Safety Assessment (PSA) are presented in this report, as the study of the system description, construction of event trees and fault trees, and the calculation of overall unavailability of the systems. It is also presented the use of microcomputer in performing some tasks, highlightning the main characteristics of a software to perform adequately the job. A sample case of fault tree construction and calculation is presented, using the PSAPACK software, distributed by the IAEA (International Atomic Energy Agency) for training purpose. (author)

  17. Bayesian-network-based safety risk analysis in construction projects

    International Nuclear Information System (INIS)

    Zhang, Limao; Wu, Xianguo; Skibniewski, Miroslaw J.; Zhong, Jingbing; Lu, Yujie

    2014-01-01

    This paper presents a systemic decision support approach for safety risk analysis under uncertainty in tunnel construction. Fuzzy Bayesian Networks (FBN) is used to investigate causal relationships between tunnel-induced damage and its influential variables based upon the risk/hazard mechanism analysis. Aiming to overcome limitations on the current probability estimation, an expert confidence indicator is proposed to ensure the reliability of the surveyed data for fuzzy probability assessment of basic risk factors. A detailed fuzzy-based inference procedure is developed, which has a capacity of implementing deductive reasoning, sensitivity analysis and abductive reasoning. The “3σ criterion” is adopted to calculate the characteristic values of a triangular fuzzy number in the probability fuzzification process, and the α-weighted valuation method is adopted for defuzzification. The construction safety analysis progress is extended to the entire life cycle of risk-prone events, including the pre-accident, during-construction continuous and post-accident control. A typical hazard concerning the tunnel leakage in the construction of Wuhan Yangtze Metro Tunnel in China is presented as a case study, in order to verify the applicability of the proposed approach. The results demonstrate the feasibility of the proposed approach and its application potential. A comparison of advantages and disadvantages between FBN and fuzzy fault tree analysis (FFTA) as risk analysis tools is also conducted. The proposed approach can be used to provide guidelines for safety analysis and management in construction projects, and thus increase the likelihood of a successful project in a complex environment. - Highlights: • A systemic Bayesian network based approach for safety risk analysis is developed. • An expert confidence indicator for probability fuzzification is proposed. • Safety risk analysis progress is extended to entire life cycle of risk-prone events. • A typical

  18. The Decision Making Trial and Evaluation Laboratory (Dematel) and Analytic Network Process (ANP) for Safety Management System Evaluation Performance

    Science.gov (United States)

    Rolita, Lisa; Surarso, Bayu; Gernowo, Rahmat

    2018-02-01

    In order to improve airport safety management system (SMS) performance, an evaluation system is required to improve on current shortcomings and maximize safety. This study suggests the integration of the DEMATEL and ANP methods in decision making processes by analyzing causal relations between the relevant criteria and taking effective analysis-based decision. The DEMATEL method builds on the ANP method in identifying the interdependencies between criteria. The input data consists of questionnaire data obtained online and then stored in an online database. Furthermore, the questionnaire data is processed using DEMATEL and ANP methods to obtain the results of determining the relationship between criteria and criteria that need to be evaluated. The study cases on this evaluation system were Adi Sutjipto International Airport, Yogyakarta (JOG); Ahmad Yani International Airport, Semarang (SRG); and Adi Sumarmo International Airport, Surakarta (SOC). The integration grades SMS performance criterion weights in a descending order as follow: safety and destination policy, safety risk management, healthcare, and safety awareness. Sturges' formula classified the results into nine grades. JOG and SMG airports were in grade 8, while SOG airport was in grade 7.

  19. Procurement in the Nuclear Industry, Quality, Safety and Decision Making; Upphandling inom kaernkraftsindustrin, kvalitet, saekerhet och beslutsfattande

    Energy Technology Data Exchange (ETDEWEB)

    Jakobsson, Marianne; Svenson, Ola; Salo, Ilkka (PSA Psykologi och beslutsfattande, Stockholm (Sweden))

    2010-03-15

    The major purpose of the present study is partly to map and partly to make an analysis of the decision processes in the procurement routines in the nuclear industry in order to provide a basis for: 1. further development of safety inspections about procurements for Swedish Radiation Safety Authority 2. improvements of safety management in connection with procurement within a nuclear-power plant, 3 improvements of procurement routines in general in a nuclear power plant. The procurement processes at a nuclear power plant were analyzed from a decision theoretic perspective. Key staff at the plant was interviewed and written instructions as well as digitalized processes were used in the analysis. The results illustrate the most important moments during the procurement process with descriptions from interviews and documents. The staff at the nuclear power plant used a multi-attribute utility decision theory MAUT-inspired model in evaluation of alternatives and both compensatory (in which negative aspects can be compensated by positive aspects) and non-compensatory (in which certain 'pass' levels of attributes have to be exceeded for a choice) decision rules were used in the procurement process. Not surprising, nuclear safety was evaluated in a non-compensatory manner following regulatory criteria while costs were evaluated in trade-off compensatory rules, which means that a weakness in one consideration might be compensated by strength in another consideration. Thus, nuclear safety above the regulator's and law requirements are not integrated in a compensatory manner when procurement alternatives are evaluated. The nuclear plant assessed an organization's safety culture at an early stage of the purchasing process. A successful and a less successful procurement case were reported with the lessons learned from them. We find that the existing written instructions for purchase were well elaborated and adequate. There is a lack of personal resources when

  20. A Semantic Approach with Decision Support for Safety Service in Smart Home Management.

    Science.gov (United States)

    Huang, Xiaoci; Yi, Jianjun; Zhu, Xiaomin; Chen, Shaoli

    2016-08-03

    Research on smart homes (SHs) has increased significantly in recent years because of the convenience provided by having an assisted living environment. The functions of SHs as mentioned in previous studies, particularly safety services, are seldom discussed or mentioned. Thus, this study proposes a semantic approach with decision support for safety service in SH management. The focus of this contribution is to explore a context awareness and reasoning approach for risk recognition in SH that enables the proper decision support for flexible safety service provision. The framework of SH based on a wireless sensor network is described from the perspective of neighbourhood management. This approach is based on the integration of semantic knowledge in which a reasoner can make decisions about risk recognition and safety service. We present a management ontology for a SH and relevant monitoring contextual information, which considers its suitability in a pervasive computing environment and is service-oriented. We also propose a rule-based reasoning method to provide decision support through reasoning techniques and context-awareness. A system prototype is developed to evaluate the feasibility, time response and extendibility of the approach. The evaluation of our approach shows that it is more effective in daily risk event recognition. The decisions for service provision are shown to be accurate.

  1. A Semantic Approach with Decision Support for Safety Service in Smart Home Management

    Directory of Open Access Journals (Sweden)

    Xiaoci Huang

    2016-08-01

    Full Text Available Research on smart homes (SHs has increased significantly in recent years because of the convenience provided by having an assisted living environment. The functions of SHs as mentioned in previous studies, particularly safety services, are seldom discussed or mentioned. Thus, this study proposes a semantic approach with decision support for safety service in SH management. The focus of this contribution is to explore a context awareness and reasoning approach for risk recognition in SH that enables the proper decision support for flexible safety service provision. The framework of SH based on a wireless sensor network is described from the perspective of neighbourhood management. This approach is based on the integration of semantic knowledge in which a reasoner can make decisions about risk recognition and safety service. We present a management ontology for a SH and relevant monitoring contextual information, which considers its suitability in a pervasive computing environment and is service-oriented. We also propose a rule-based reasoning method to provide decision support through reasoning techniques and context-awareness. A system prototype is developed to evaluate the feasibility, time response and extendibility of the approach. The evaluation of our approach shows that it is more effective in daily risk event recognition. The decisions for service provision are shown to be accurate.

  2. Operation and safety decision-making support expert system in NPP

    International Nuclear Information System (INIS)

    Wei Yanhui; Su Desong; Chen Weihua; Zhang Jianbo

    2014-01-01

    The article first reviewed three operation support systems currently used in NPP: real-time information surveillance system, important equipment surveillance system and plant process control and monitoring system, then presents the structure and function of three expert support sub-systems (intelligent alarm monitoring system, computer-based operating procedure support system, safety information expert decision support system). Finally the article discussed the meaning of a kind of operation decision making support system. (authors)

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  4. A Weibull Approach for Enabling Safety-Oriented Decision-Making for Electronic Railway Signaling Systems

    Directory of Open Access Journals (Sweden)

    Emanuele Pascale

    2018-04-01

    Full Text Available This paper presents the advantages of using Weibull distributions, within the context of railway signaling systems, for enabling safety-oriented decision-making. Failure rates are used to statistically model the basic event of fault-tree analysis, and their value sizes the maximum allowable latency of failures to fulfill the safety target for which the system has been designed. Relying on field-return failure data, Weibull parameters have been calculated for an existing electronic signaling system and a comparison with existing predictive reliability data, based on exponential distribution, is provided. Results are discussed in order to drive considerations on the respect of quantitative targets and on the impact that a wrong hypothesis might have on the choice of a given architecture. Despite the huge amount of information gathered through the after-sales logbook used to build reliability distribution, several key elements for reliable estimation of failure rate values are still missing. This might affect the uncertainty of reliability parameters and the effort required to collect all the information. We then present how to intervene when operational failure rates present higher values compared to the theoretical approach: increasing the redundancies of the system or performing preventive maintenance tasks. Possible consequences of unjustified adoption of constant failure rate are presented. Some recommendations are also shared in order to build reliability-oriented logbooks and avoid data censoring phenomena by enhancing the functions of the electronic boards composing the system.

  5. True pancreaticoduodenal artery aneurysms: A decision analysis

    International Nuclear Information System (INIS)

    Takao, Hidemasa; Nojo, Takeshi; Ohtomo, Kuni

    2010-01-01

    Purpose: True pancreaticoduodenal artery aneurysms are rare. No definitive study evaluating the natural history of these lesions or their preferred method of treatment has been published. The purpose of this study was to evaluate the outcome of preventive treatment of unruptured pancreaticoduodenal artery aneurysms using a Markov model. Materials and methods: With the use of a Markov model, we performed a decision analysis to evaluate the outcome of preventive treatment of unruptured pancreaticoduodenal artery aneurysms. The risk of rupture and the mortality of preventive treatment are unknown. Therefore, we performed sensitivity analysis using these parameters. Effectiveness was measured in life expectancy. Results: For 80-year-old patients, preventive treatment was dominated by no treatment if mortality rates of preventive treatment were greater than 1.4%, greater than 2.6%, greater than 3.8%, and greater than 4.8% at annual rupture rates of 1%, 2%, 3%, and 4%, respectively. For 50-year-old patients, preventive treatment was dominated by no treatment if mortality rates of preventive treatment were greater than 3.3%, greater than 5.9%, greater than 8.0%, and greater than 9.7% at annual rupture rates of 1%, 2%, 3%, and 4%, respectively. Conclusion: The effectiveness of preventive treatment of unruptured pancreaticoduodenal artery aneurysms depends on the aneurysm rupture rate, mortality rate of preventive treatment, and patient age. Taking into account the effects of these parameters is important in making treatment decisions.

  6. Classification analysis of organization factors related to system safety

    International Nuclear Information System (INIS)

    Liu Huizhen; Zhang Li; Zhang Yuling; Guan Shihua

    2009-01-01

    This paper analyzes the different types of organization factors which influence the system safety. The organization factor can be divided into the interior organization factor and exterior organization factor. The latter includes the factors of political, economical, technical, law, social culture and geographical, and the relationships among different interest groups. The former includes organization culture, communication, decision, training, process, supervision and management and organization structure. This paper focuses on the description of the organization factors. The classification analysis of the organization factors is the early work of quantitative analysis. (authors)

  7. Decisions under uncertainty using Bayesian analysis

    Directory of Open Access Journals (Sweden)

    Stelian STANCU

    2006-01-01

    Full Text Available The present paper makes a short presentation of the Bayesian decions method, where extrainformation brings a great support to decision making process, but also attract new costs. In this situation, getting new information, generally experimentaly based, contributes to diminushing the uncertainty degree that influences decision making process. As a conclusion, in a large number of decision problems, there is the possibility that the decision makers will renew some decisions already taken because of the facilities offered by obtainig extrainformation.

  8. Reliability analysis of Angra I safety systems

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

    1980-07-01

    An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

  9. Development of evaluation method for software safety analysis techniques

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. Analysis and design on airport safety information management system

    Directory of Open Access Journals (Sweden)

    Yan Lin

    2017-01-01

    Full Text Available Airport safety information management system is the foundation of implementing safety operation, risk control, safety performance monitor, and safety management decision for the airport. The paper puts forward the architecture of airport safety information management system based on B/S model, focuses on safety information processing flow, designs the functional modules and proposes the supporting conditions for system operation. The system construction is helpful to perfecting the long effect mechanism driven by safety information, continually increasing airport safety management level and control proficiency.

  11. Application of Software Safety Analysis Methods

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  12. From Safety Analysis to Formal Specification

    DEFF Research Database (Denmark)

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

    1998-01-01

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

  13. Swimming pool reactor reliability and safety analysis

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1997-01-01

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

  14. Robustness of Multiple Objective Decision Analysis Preference Functions

    Science.gov (United States)

    2002-06-01

    Bayesian Decision Theory and Utilitarian Ethics ,” American Economic Review Papers and Proceedings, 68: 223-228 (May 1978). Hartsough, Bruce R. “A...1983). Morrell, Darryl and Eric Driver. “ Bayesian Network Implementation of Levi’s Epistemic Utility Decision Theory ,” International Journal Of...elicitation efficiency for the decision maker. Subject Terms Decision Analysis, Utility Theory , Elicitation Error, Operations Research, Decision

  15. On the Safety of Machine Learning: Cyber-Physical Systems, Decision Sciences, and Data Products.

    Science.gov (United States)

    Varshney, Kush R; Alemzadeh, Homa

    2017-09-01

    Machine learning algorithms increasingly influence our decisions and interact with us in all parts of our daily lives. Therefore, just as we consider the safety of power plants, highways, and a variety of other engineered socio-technical systems, we must also take into account the safety of systems involving machine learning. Heretofore, the definition of safety has not been formalized in a machine learning context. In this article, we do so by defining machine learning safety in terms of risk, epistemic uncertainty, and the harm incurred by unwanted outcomes. We then use this definition to examine safety in all sorts of applications in cyber-physical systems, decision sciences, and data products. We find that the foundational principle of modern statistical machine learning, empirical risk minimization, is not always a sufficient objective. We discuss how four different categories of strategies for achieving safety in engineering, including inherently safe design, safety reserves, safe fail, and procedural safeguards can be mapped to a machine learning context. We then discuss example techniques that can be adopted in each category, such as considering interpretability and causality of predictive models, objective functions beyond expected prediction accuracy, human involvement for labeling difficult or rare examples, and user experience design of software and open data.

  16. Analysis of Aviation Safety Reporting System Incident Data Associated With the Technical Challenges of the Vehicle Systems Safety Technology Project

    Science.gov (United States)

    Withrow, Colleen A.; Reveley, Mary S.

    2014-01-01

    This analysis was conducted to support the Vehicle Systems Safety Technology (VSST) Project of the Aviation Safety Program (AVsP) milestone VSST4.2.1.01, "Identification of VSST-Related Trends." In particular, this is a review of incident data from the NASA Aviation Safety Reporting System (ASRS). The following three VSST-related technical challenges (TCs) were the focus of the incidents searched in the ASRS database: (1) Vechicle health assurance, (2) Effective crew-system interactions and decisions in all conditions; and (3) Aircraft loss of control prevention, mitigation, and recovery.

  17. Application of Mixed Group Decision Making to Safety Evaluation of Agricultural Products

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    In view of the gravity of issues concerning safety of agricultural products and urgency of resolving these issues,after analyzing the problems existing in safety of agricultural products,this article offers a method for evaluating safety of agricultural products on the basis of mixed group decision making.First of all,it introduces the factors influencing safety evaluation of agricultural products;subsequently,given that the judgment matrices offered by the group of experts contain both reciprocal and complementary judgment matrices in the process of jointly participating in evaluation arising from personal preference,it proposes to assemble expert information in order to obtain indicator weight using the OWA operator;finally,the process of evaluating safety of agricultural products is given.

  18. Environmental sustainable decision making – The need and obstacles for integration of LCA into decision analysis

    DEFF Research Database (Denmark)

    Dong, Yan; Miraglia, Simona; Manzo, Stefano

    2018-01-01

    systems, revealing potential problem shifting between life cycle stages. Through the integration with traditional risk based decision analysis, LCA may thus facilitate a better informed decision process. In this study we explore how environmental impacts are taken into account in different fields......Decision analysis is often used to help decision makers choose among alternatives, based on the expected utility associated to each alternative as function of its consequences and potential impacts. Environmental impacts are not always among the prioritized concerns of traditional decision making...... of interest for decision makers to identify the need, potential and obstacles for integrating LCA into conventional approaches to decision problems. Three application areas are used as examples: transportation planning, flood management, and food production and consumption. The analysis of these cases shows...

  19. Using discriminant analysis for credit decision

    Directory of Open Access Journals (Sweden)

    Gheorghiţa DINCĂ

    2015-12-01

    Full Text Available This paper follows to highlight the link between the results obtained applying discriminant analysis and lending decision. For this purpose, we have carried out the research on a sample of 24 Romanian private companies, pertaining to 12 different economic sectors, from I and II categories of Bucharest Stock Exchange, for the period 2010-2012. Our study works with two popular bankruptcy risk’s prediction models, the Altman model and the Anghel model. We have double-checked and confirmed the results of our research by comparing the results from applying the two fore-mentioned models as well as by checking existing debt commitments of each analyzed company to credit institutions during the 2010-2012 period. The aim of this paper was the classification of studied companies into potential bankrupt and non-bankrupt, to assist credit institutions in their decision to grant credit, understanding the approval or rejection algorithm of loan applications and even help potential investors in these ompanies.

  20. Use of decision criteria based on expected values to support decision-making in a production assurance and safety setting

    International Nuclear Information System (INIS)

    Aven, T.; Flage, R.

    2009-01-01

    We consider decision problems related to production assurance and safety. The issue is to what extent we should use decision criteria based on expected values, such as the expected net present value (E[NPV]) and the expected cost per expected number of saved lives (ICAF), to guide the decision. Such criteria are recognised as practical tools for supporting decision-making under uncertainty, but is uncertainty adequately taken into account by these criteria? Based on the prevailing practice and the existing literature, we conclude that there is a need for a clarification of the rationale of these criteria. Adjustments of the standard approaches have been suggested to reflect risks and uncertainties, but can cautionary and precautionary concerns be replaced by formulae and mechanical procedures? These issues are discussed in the present paper, particularly addressing the company level. We argue that the search for such formulae and procedures should be replaced by a more balanced perspective acknowledging that there will always be a need for management review and judgment beyond the realm of the analyses. Most of the suggested adjustments of the E[NPV] and ICAF approaches should be avoided. They add more confusion than value.

  1. Acid rain compliance planning using decision analysis

    International Nuclear Information System (INIS)

    Norris, C.; Sweet, T.; Borison, A.

    1991-01-01

    Illinois Power Company (IP) is an investor-owned electric and natural gas utility serving portions of downstate Illinois. In addition to one nuclear unit and several small gas and/or oil-fired units, IP has ten coal-fired units. It is easy to understand the impact the Clean Air Act Amendments of 1990 (CAAA) could have on IP. Prior to passage of the CAAA, IP formed several teams to evaluate the specific compliance options at each of the high sulfur coal units. Following that effort, numerous economic analyses of compliance strategies were conducted. The CAAA have introduced a new dimension to planning under uncertainty. Not only are many of the familiar variables uncertain, but the specific form of regulation, and indeed, the compliance goal itself is hard to define. For IP, this led them to use techniques not widely used within their corporation. This paper summarizes the analytical methods used in these analyses and the preliminary results as of July, 1991. The analysis used three approaches to examine the acid rain compliance decision. These approaches were: (1) the 'most-likely,' or single-path scenario approach; (2) a multi-path strategy analysis using the strategies defined in the single-scenario analysis; and (3) a less constrained multi-path option analysis which selects the least cost compliance option for each unit

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

    International Nuclear Information System (INIS)

    COLLOPY, M.T.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Tsuchiya, Mitsuo; Kamimae, Rie

    1999-01-01

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

  4. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  5. Variable precision rough set for multiple decision attribute analysis

    Institute of Scientific and Technical Information of China (English)

    Lai; Kin; Keung

    2008-01-01

    A variable precision rough set (VPRS) model is used to solve the multi-attribute decision analysis (MADA) problem with multiple conflicting decision attributes and multiple condition attributes. By introducing confidence measures and a β-reduct, the VPRS model can rationally solve the conflicting decision analysis problem with multiple decision attributes and multiple condition attributes. For illustration, a medical diagnosis example is utilized to show the feasibility of the VPRS model in solving the MADA...

  6. Stress influences decisions to break a safety rule in a complex simulation task in females.

    Science.gov (United States)

    Starcke, Katrin; Brand, Matthias; Kluge, Annette

    2016-07-01

    The current study examines the effects of acutely induced laboratory stress on a complex decision-making task, the Waste Water Treatment Simulation. Participants are instructed to follow a certain decision rule according to safety guidelines. Violations of this rule are associated with potential high rewards (working faster and earning more money) but also with the risk of a catastrophe (an explosion). Stress was induced with the Trier Social Stress Test while control participants underwent a non-stress condition. In the simulation task, stressed females broke the safety rule more often than unstressed females: χ(2) (1, N=24)=10.36, pbreak the safety rule because stressed female participants focused on the potential high gains while they neglected the risk of potential negative consequences. Copyright © 2016 Elsevier B.V. All rights reserved.

  7. A decision model to allocate protective safety barriers and mitigate domino effects

    International Nuclear Information System (INIS)

    Janssens, Jochen; Talarico, Luca; Reniers, Genserik; Sörensen, Kenneth

    2015-01-01

    In this paper, we present a model to support decision-makers about where to locate safety barriers and mitigate the consequences of an accident triggering domino effects. Based on the features of an industrial area that may be affected by domino accidents, and knowing the characteristics of the safety barriers that can be installed to stall the fire propagation between installations, the decision model can help practitioners in their decision-making. The model can be effectively used to decide how to allocate a limited budget in terms of safety barriers. The goal is to maximize the time-to-failure of a chemical installation ensuring a worst case scenario approach. The model is mathematically stated and a flexible and effective solution approach, based on metaheuristics, is developed and tested on an illustrative case study representing a tank storage area of a chemical company. We show that a myopic optimization approach, which does not take into account knock-on effects possibly triggered by an accident, can lead to a distribution of safety barriers that are not effective in mitigating the consequences of a domino accident. Moreover, the optimal allocation of safety barriers, when domino effects are considered, may depend on the so-called cardinality of the domino effects. - Highlights: • A model to allocate safety barriers and mitigate domino effects is proposed. • The goal is to maximize the escalation time of the worst case scenario. • The model provides useful recommendations for decision makers. • A fast metaheuristic approach is proposed to solve such a complex problem. • Numerical simulations on a realistic case study are shown

  8. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    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

  9. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Assumptions and Policy Decisions for Vital Area Identification Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myungsu; Bae, Yeon-Kyoung; Lee, Youngseung [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    U.S. Nuclear Regulatory Commission and IAEA guidance indicate that certain assumptions and policy questions should be addressed to a Vital Area Identification (VAI) process. Korea Hydro and Nuclear Power conducted a VAI based on current Design Basis Threat and engineering judgement to identify APR1400 vital areas. Some of the assumptions were inherited from Probabilistic Safety Assessment (PSA) as a sabotage logic model was based on PSA logic tree and equipment location data. This paper illustrates some important assumptions and policy decisions for APR1400 VAI analysis. Assumptions and policy decisions could be overlooked at the beginning stage of VAI, however they should be carefully reviewed and discussed among engineers, plant operators, and regulators. Through APR1400 VAI process, some of the policy concerns and assumptions for analysis were applied based on document research and expert panel discussions. It was also found that there are more assumptions to define for further studies for other types of nuclear power plants. One of the assumptions is mission time, which was inherited from PSA.

  11. Fault trees for decision making in systems analysis

    International Nuclear Information System (INIS)

    Lambert, H.E.

    1975-01-01

    The application of fault tree analysis (FTA) to system safety and reliability is presented within the framework of system safety analysis. The concepts and techniques involved in manual and automated fault tree construction are described and their differences noted. The theory of mathematical reliability pertinent to FTA is presented with emphasis on engineering applications. An outline of the quantitative reliability techniques of the Reactor Safety Study is given. Concepts of probabilistic importance are presented within the fault tree framework and applied to the areas of system design, diagnosis and simulation. The computer code IMPORTANCE ranks basic events and cut sets according to a sensitivity analysis. A useful feature of the IMPORTANCE code is that it can accept relative failure data as input. The output of the IMPORTANCE code can assist an analyst in finding weaknesses in system design and operation, suggest the most optimal course of system upgrade, and determine the optimal location of sensors within a system. A general simulation model of system failure in terms of fault tree logic is described. The model is intended for efficient diagnosis of the causes of system failure in the event of a system breakdown. It can also be used to assist an operator in making decisions under a time constraint regarding the future course of operations. The model is well suited for computer implementation. New results incorporated in the simulation model include an algorithm to generate repair checklists on the basis of fault tree logic and a one-step-ahead optimization procedure that minimizes the expected time to diagnose system failure. (80 figures, 20 tables)

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

    Energy Technology Data Exchange (ETDEWEB)

    Delage, M; Giroux, C

    1990-12-01

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

  13. Accident Analysis and Highway Safety

    Directory of Open Access Journals (Sweden)

    Omar Noorliyana

    2017-01-01

    Full Text Available Since 2010, Federal Route FT050 (Jalan Batu Pahat-Kluang has undergone many changes, including the improvement of geometric features (i.e., construction of median, dedicated U-turns and additional lanes and upgrading the quality of the road surface. Unfortunately, even with these enhancements, accidents continue to occur along this route. This study covered both accident analysis and blackspot study. Accident point weightage was used to identify blackspot locations. The results reveal hazardous road locations and blackspot ranking along the route.

  14. Safety analysis methodology for OPR 1000

    International Nuclear Information System (INIS)

    Hwang-Yong, Jun

    2005-01-01

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

  15. Modelling elderly cardiac patients decision making using Cognitive Work Analysis: identifying requirements for patient decision aids.

    Science.gov (United States)

    Dhukaram, Anandhi Vivekanandan; Baber, Chris

    2015-06-01

    Patients make various healthcare decisions on a daily basis. Such day-to-day decision making can have significant consequences on their own health, treatment, care, and costs. While decision aids (DAs) provide effective support in enhancing patient's decision making, to date there have been few studies examining patient's decision making process or exploring how the understanding of such decision processes can aid in extracting requirements for the design of DAs. This paper applies Cognitive Work Analysis (CWA) to analyse patient's decision making in order to inform requirements for supporting self-care decision making. This study uses focus groups to elicit information from elderly cardiovascular disease (CVD) patients concerning a range of decision situations they face on a daily basis. Specifically, the focus groups addressed issues related to the decision making of CVD in terms of medication compliance, pain, diet and exercise. The results of these focus groups are used to develop high level views using CWA. CWA framework decomposes the complex decision making problem to inform three approaches to DA design: one design based on high level requirements; one based on a normative model of decision-making for patients; and the third based on a range of heuristics that patients seem to use. CWA helps in extracting and synthesising decision making from different perspectives: decision processes, work organisation, patient competencies and strategies used in decision making. As decision making can be influenced by human behaviour like skills, rules and knowledge, it is argued that patients require support to different types of decision making. This paper also provides insights for designers in using CWA framework for the design of effective DAs to support patients in self-management. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  16. Reliability analysis of software based safety functions

    International Nuclear Information System (INIS)

    Pulkkinen, U.

    1993-05-01

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

  17. Decision making with epistemic uncertainty under safety constraints: An application to seismic design

    Science.gov (United States)

    Veneziano, D.; Agarwal, A.; Karaca, E.

    2009-01-01

    The problem of accounting for epistemic uncertainty in risk management decisions is conceptually straightforward, but is riddled with practical difficulties. Simple approximations are often used whereby future variations in epistemic uncertainty are ignored or worst-case scenarios are postulated. These strategies tend to produce sub-optimal decisions. We develop a general framework based on Bayesian decision theory and exemplify it for the case of seismic design of buildings. When temporal fluctuations of the epistemic uncertainties and regulatory safety constraints are included, the optimal level of seismic protection exceeds the normative level at the time of construction. Optimal Bayesian decisions do not depend on the aleatory or epistemic nature of the uncertainties, but only on the total (epistemic plus aleatory) uncertainty and how that total uncertainty varies randomly during the lifetime of the project. ?? 2009 Elsevier Ltd. All rights reserved.

  18. K West integrated water treatment system subproject safety analysis document

    International Nuclear Information System (INIS)

    SEMMENS, L.S.

    1999-01-01

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System

  19. K West integrated water treatment system subproject safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    SEMMENS, L.S.

    1999-02-24

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

  20. CRITICAL ANALYSIS OF THE RELIABILITY OF INTUITIVE MORAL DECISIONS

    Directory of Open Access Journals (Sweden)

    V. V. Nadurak

    2017-06-01

    Full Text Available Purpose of the research is a critical analysis of the reliability of intuitive moral decisions. Methodology. The work is based on the methodological attitude of empirical ethics, involving the use of findings from empirical research in ethical reflection and decision making. Originality. The main kinds of intuitive moral decisions are identified: 1 intuitively emotional decisions (i.e. decisions made under the influence of emotions that accompanies the process of moral decision making; 2 decisions made under the influence of moral risky psychological aptitudes (unconscious human tendencies that makes us think in a certain way and make decisions, unacceptable from the logical and ethical point of view; 3 intuitively normative decisions (decisions made under the influence of socially learned norms, that cause evaluative feeling «good-bad», without conscious reasoning. It was found that all of these kinds of intuitive moral decisions can lead to mistakes in the moral life. Conclusions. Considering the fact that intuition systematically leads to erroneous moral decisions, intuitive reaction cannot be the only source for making such decisions. The conscious rational reasoning can compensate for weaknesses of intuition. In this case, there is a necessity in theoretical model that would structure the knowledge about the interactions between intuitive and rational factors in moral decisions making and became the basis for making suggestions that would help us to make the right moral decision.

  1. Preliminary safety analysis report for the TFTR

    International Nuclear Information System (INIS)

    Lind, K.E.; Levine, J.D.; Howe, H.J.

    A Preliminary Safety Analysis Report has been prepared for the Tokamak Fusion Test Reactor. No accident scenarios have been identified which would result in exposures to on-site personnel or the general public in excess of the guidelines defined for the project by DOE

  2. Safety analysis of accident localization system

    International Nuclear Information System (INIS)

    1999-01-01

    A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

  3. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

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

  4. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  5. Integrated framework for dynamic safety analysis

    International Nuclear Information System (INIS)

    Kim, Tae Wan; Karanki, Durga R.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    1989-07-01

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

  7. DECISION ANALYSIS OF INCINERATION COSTS IN SUPERFUND SITE REMEDIATION

    Science.gov (United States)

    This study examines the decision-making process of the remedial design (RD) phase of on-site incineration projects conducted at Superfund sites. Decisions made during RD affect the cost and schedule of remedial action (RA). Decision analysis techniques are used to determine the...

  8. Uncertainty analysis for Ulysses safety evaluation report

    International Nuclear Information System (INIS)

    Frank, M.V.

    1991-01-01

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

  9. HSE assessment of explosion risk analysis in offshore safety cases

    Energy Technology Data Exchange (ETDEWEB)

    Brighton, P.W.M.; Fearnley, P.J.; Brearley, I.G. [Health and Safety Executive, Bootle (United Kingdom). Offshore Safety Div.

    1995-12-31

    In the past two years HSE has assessed around 250 Safety Cases for offshore oil and gas installations, building up a unique overview of the current state of the art on fire and explosion risk assessment. This paper reviews the explosion risk methods employed, focusing on the aspects causing most difficulty for assessment and acceptance of Safety Cases. Prediction of overpressures in offshore explosions has been intensively researched in recent years but the justification of the means of prevention, control and mitigation of explosions often depends on much additional analysis of the frequency and damage potential of explosions. This involves a number of factors, the five usually considered being: leak sizes; gas dispersion; ignition probabilities; the frequency distribution of explosion strength; and the prediction of explosion damage. Sources of major uncertainty in these factors and their implications for practical risk management decisions are discussed. (author)

  10. Safety assessment of Olkiluoto NPP units 1 and 2. Decision of the Radiation and Nuclear Safety Authority regarding the periodic safety review of the Olkiluoto NPP

    International Nuclear Information System (INIS)

    2010-02-01

    In this safety assessment the Radiation and Nuclear Safety Authority (STUK) has evaluated the safety of the Olkiluoto Nuclear Power Plant units 1 and 2 in connection with the periodic safety review. This safety assessment provides a summary of the reviews, inspections and continuous oversight carried out by STUK. The issues addressed in the assessment and the related evaluation criteria are set forth in the nuclear energy and radiation safety legislation and the regulations issued thereunder. The provisions of the Nuclear Energy Act concerning the safe use of nuclear energy, security and emergency preparedness arrangements, and waste management are specified in more detail in the Government Decrees and Regulatory Guides issued by STUK. Based on the assessment, STUK consideres that the Olkiluoto Nuclear Power Plant units 1 and 2 meet the set safety requirements for operational nuclear power plants, the emergency preparedness arrangements are sufficient and the necessary control to prevent the proliferation of nuclear weapons has been appropriately arranged. The physical protection of the Olkiluoto nuclear power plant is not yet completely in compliance with the requirements of Government Decree 734/2008, which came into force in December 2008. Further requirements concerning this issue based also on the principle of continuous improvement were included in the decision relating to the periodic safety review. The safety of the Olkiluoto nuclear power plant was assessed in compliance with the Government Decree on the Safety of Nuclear Power Plants (733/2008), which came into force in 2008. The decree notes that existing nuclear power plants need not meet all the requirements set out for new plants. Most of the design bases pertaining to the Olkiluoto 1 and 2 nuclear power plant units were set in the 1970s. Substantial modernisations have been carried out at the Olkiluoto 1 and 2 nuclear power plant units since their commissioning to improve safety. This is in line with

  11. Deterministic and probabilistic approach to safety analysis

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1980-01-01

    The examples discussed in this paper show that reliability analysis methods fairly well can be applied in order to interpret deterministic safety criteria in quantitative terms. For further improved extension of applied reliability analysis it has turned out that the influence of operational and control systems and of component protection devices should be considered with the aid of reliability analysis methods in detail. Of course, an extension of probabilistic analysis must be accompanied by further development of the methods and a broadening of the data base. (orig.)

  12. Integrating risk management and safety culture in a framework for risk informed decision making

    International Nuclear Information System (INIS)

    Nelson, W.R.

    2009-01-01

    Operators and regulators of nuclear power plants agree on the importance of maintaining safety and controlling accident risks. Effective safety and risk management requires treatment of both technical and organizational components. Probabilistic Risk Assessment (PRA) provides tools for technical risk management. However, organizational factors are not treated in PRA, but are addressed using different approaches. To bring both components together, a framework of Risk Informed Decision Making (RIDM) is needed. The objective tree structure of the International Atomic Energy Agency (IAEA) is a promising approach to combine both elements. Effective collaboration involving regulatory and industry groups is needed to accomplish the integration. (author)

  13. Decision analysis in the clinical neurosciences

    NARCIS (Netherlands)

    D.W.J. Dippel (Diederik)

    1994-01-01

    textabstractDiagnostic and therapeutic choice in neurology can fortunately be made without formal decision support in the majority of cases. in many patients a diagnosis and treatment choice are relatively easy to establish. This study however, concerns the application of a decision support

  14. Qualitative uncertainty analysis in probabilistic safety assessment context

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

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

  16. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  17. The PEC reactor. Safety analysis: Detailed reports

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

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

  18. Impact of support system failure limitations on probabilistic safety assessment and in regulatory decision making

    International Nuclear Information System (INIS)

    Bickel, J.H.

    1990-01-01

    When used as a tool for safety decision making, Probabilistic Safety Assessment (PSA) is as effective as it realistically characterizes the overall frequency and consequences of various types of system and component failures. If significant support system failure events are omitted from consideration, the PSA process omits the characterization of possible unique contributors to core damage risk, possibly underestimates the frequency of core damage, and reduces the future utility of the PSA as a decision making tool for the omitted support system. This paper is based on a review of several recent US PSA studies and the author's participation in several International Atomic Energy Agency (IAEA) sponsored peer reviews. 21 refs., 2 figs., 1 tab

  19. Environmental Decision Analysis: Meeting the Challenges of Making Good Decisions at CALFED

    Directory of Open Access Journals (Sweden)

    Claire D Tomkins

    2006-09-01

    Full Text Available We present a methodology to support decision making at CALFED based on the principles of decision analysis, an analytical approach to decision making designed to handle complex decisions involving both uncertainty and multiple dimensions of value. The impetus for such an approach is a recognized need to enhance communication between scientists and management and between program elements within CALFED. In addition, the environmental decision analysis framework supports both the explicit representation of uncertainty in the decision problem and communication about risk, important elements of most environmental management decisions. The decision analysis cycle consists of four phases: 1 formulate, 2 evaluate, 3 appraise, and 4 decide. In phase one, we identify the objectives and also the alternatives, or possible actions. To facilitate inter-comparison between proposed actions, we recommend formulation of a set of common metrics for CALFED. In our pilot study, we introduced common metrics for salinity, winter-run Chinook salmon survival, and habitat health. The second phase focuses on quantifying possible impacts on the set of metrics, drawing on existing data, model runs, and expert opinions. For the evaluation phase, we employ tools such as decision trees to assess the system-wide impacts of a given action. In the final phase, tools such as expected cost-benefit analysis, value contribution diagrams, and 3-D tradeoff plots aid communication between various stakeholders, scientists, and managers. While decision analysis provides a spectrum of decision support tools, we emphasize that it does not dictate a solution but rather enhances communication about tradeoffs associated with different actions.

  20. Risk-based decision analysis for groundwater operable units

    International Nuclear Information System (INIS)

    Chiaramonte, G.R.

    1995-01-01

    This document proposes a streamlined approach and methodology for performing risk assessment in support of interim remedial measure (IRM) decisions involving the remediation of contaminated groundwater on the Hanford Site. This methodology, referred to as ''risk-based decision analysis,'' also supports the specification of target cleanup volumes and provides a basis for design and operation of the groundwater remedies. The risk-based decision analysis can be completed within a short time frame and concisely documented. The risk-based decision analysis is more versatile than the qualitative risk assessment (QRA), because it not only supports the need for IRMs, but also provides criteria for defining the success of the IRMs and provides the risk-basis for decisions on final remedies. For these reasons, it is proposed that, for groundwater operable units, the risk-based decision analysis should replace the more elaborate, costly, and time-consuming QRA

  1. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

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

    2000-12-01

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

  2. Safety analysis and risk assessment handbook

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  4. LFR safety approach and main ELFR safety analysis results

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Demin, V.F.

    2002-01-01

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

  6. Info-gap decision theory decisions under severe uncertainty

    CERN Document Server

    Ben-Haim, Yakov

    2006-01-01

    Everyone makes decisions, but not everyone is a decision analyst. A decision analyst uses quantitative models and computational methods to formulate decision algorithms, assess decision performance, identify and evaluate options, determine trade-offs and risks, evaluate strategies for investigation, and so on. This book is written for decision analysts. The term ""decision analyst"" covers an extremely broad range of practitioners. Virtually all engineers involved in design (of buildings, machines, processes, etc.) or analysis (of safety, reliability, feasibility, etc.) are decision analysts,

  7. Decision Analysis and Its Application to the Frequency of Containment Integrated Leakage Rate Tests

    International Nuclear Information System (INIS)

    Apostolakis, George E.; Koser, John P.; Sato, Gaku

    2004-01-01

    For nuclear utilities to become competitive in a deregulated electricity market, costs must be reduced, safety must be maintained, and interested stakeholders must remain content with the decisions being made. One way to reduce costs is to reduce the frequency of preventive maintenance and testing. However, these changes must be weighed against their impact on safety and stakeholder relations. We present a methodology that allows the evaluation of decision options using a number of objectives that include safety, economics, and stakeholder relations. First, the candidate decision options are screened to make sure that they satisfy the relevant regulatory requirements. The remaining options are evaluated using multiattribute utility theory. The results of the formal analysis include a ranking of the options according to their desirability as well as the major reasons that explain this ranking. These results are submitted to a deliberative process in which the decision makers scrutinize the results to ensure that they are meaningful. During the deliberation, new decision options may be formulated based on the insights that the formal analysis provides, as happened in the case study of this paper. This case study deals with the reduction in frequency of the containment integrated leak rate test of a boiling water reactor

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Reinhart, M.

    2007-01-01

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

  10. Decision theory, the context for risk and reliability analysis

    International Nuclear Information System (INIS)

    Kaplan, S.

    1985-01-01

    According to this model of the decision process then, the optimum decision is that option having the largest expected utility. This is the fundamental model of a decision situation. It is necessary to remark that in order for the model to represent a real-life decision situation, it must include all the options present in that situation, including, for example, the option of not deciding--which is itself a decision, although usually not the optimum one. Similarly, it should include the option of delaying the decision while the authors gather further information. Both of these options have probabilities, outcomes, impacts, and utilities like any option and should be included explicitly in the decision diagram. The reason for doing a quantitative risk or reliability analysis is always that, somewhere underlying there is a decision to be made. The decision analysis therefore always forms the context for the risk or reliability analysis, and this context shapes the form and language of that analysis. Therefore, they give in this section a brief review of the well-known decision theory diagram

  11. COLD-SAT feasibility study safety analysis

    Science.gov (United States)

    Mchenry, Steven T.; Yost, James M.

    1991-01-01

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

  12. Safety of GM crops: compositional analysis.

    Science.gov (United States)

    Brune, Philip D; Culler, Angela Hendrickson; Ridley, William P; Walker, Kate

    2013-09-04

    The compositional analysis of genetically modified (GM) crops has continued to be an important part of the overall evaluation in the safety assessment program for these materials. The variety and complexity of genetically engineered traits and modes of action that will be used in GM crops in the near future, as well as our expanded knowledge of compositional variability and factors that can affect composition, raise questions about compositional analysis and how it should be applied to evaluate the safety of traits. The International Life Sciences Institute (ILSI), a nonprofit foundation whose mission is to provide science that improves public health and well-being by fostering collaboration among experts from academia, government, and industry, convened a workshop in September 2012 to examine these and related questions, and a series of papers has been assembled to describe the outcomes of that meeting.

  13. Qualitative safety analysis in accelerator based systems

    International Nuclear Information System (INIS)

    Sarkar, P.K.; Chowdhury, Lekha M.

    2006-01-01

    In recent developments connected to high energy and high current accelerators, the accelerator driven systems (ADS) and the Radioactive Ion Beam (RIB) facilities come in the forefront of application. For medical and industrial applications high current accelerators often need to be located in populated areas. These facilities pose significant radiological hazard during their operation and accidental situations. We have done a qualitative evaluation of radiological safety analysis using the probabilistic safety analysis (PSA) methods for accelerator-based systems. The major contribution to hazard comes from a target rupture scenario in both ADS and RIB facilities. Other significant contributors to hazard in the facilities are also discussed using fault tree and event tree methodologies. (author)

  14. Computational methods for nuclear criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.

    1992-01-01

    Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

  15. Decision forests for computer vision and medical image analysis

    CERN Document Server

    Criminisi, A

    2013-01-01

    This practical and easy-to-follow text explores the theoretical underpinnings of decision forests, organizing the vast existing literature on the field within a new, general-purpose forest model. Topics and features: with a foreword by Prof. Y. Amit and Prof. D. Geman, recounting their participation in the development of decision forests; introduces a flexible decision forest model, capable of addressing a large and diverse set of image and video analysis tasks; investigates both the theoretical foundations and the practical implementation of decision forests; discusses the use of decision for

  16. Use of the Safety Monitor in operational decision-making at a nuclear generating facility

    International Nuclear Information System (INIS)

    Chien, Shan H.; Hook, Thomas G.; Lee, Roger J.

    1998-01-01

    The utilization of Safety Monitor at a nuclear generating facility in 1994 revolutionized the way US nuclear power plants manage configuration risks. At Southern California Edison (SCE) Company's San Onofre Nuclear Generating Station, it transformed probabilistic risk assessment (PRA) from a retrospective tool for understanding past risk into a prospective tool for controlling future risk. Since that time, many other nuclear utilities have taken aggressive steps in using PRA better to understand and manage risks associated with plant operation and maintenance. These utilities have employed a variety of methods ranging from systems similar to San Onofre's Safety Monitor to systems dramatically different in both technology and philosophy. In the development and use of its Safety Monitor, SCE has been guided by two philosophical goals: (1) maximize the objectivity of PRA-informed decision-making relative to managing configuration risks, and (2) ensure that risks are managed conservatively

  17. Computer graphics in reactor safety analysis

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.

    1989-01-01

    This paper describes a family of three computer graphics codes designed to assist the analyst in three areas: the modelling of complex three-dimensional finite element models of reactor structures; the interpretation of computational results; and the reporting of the results of numerical simulations. The purpose and key features of each code are presented. The graphics output used in actual safety analysis are used to illustrate the capabilities of each code. 5 refs., 10 figs

  18. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  19. N Reactor updated safety analysis report, NUSAR

    International Nuclear Information System (INIS)

    1978-01-01

    An update of the N Reactor safety analysis is presented to reconfirm that the continued operation does not pose undue risk to DOE personnel and property, the public, or the environment. A reanalysis of LOCA and reactivity transients utilizing current codes and methods is made. The principal aspects of the overall submission, a general description, and site characteristics including geography and demography, nearby industrial, transportation and military facilities, meteorology, hydraulic engineering, and geology and seismology are described

  20. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

    Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

    1985-01-01

    The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Jamali, K.

    1997-01-01

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

  2. Latest developments on safety analysis methodologies at the Juzbado plant

    International Nuclear Information System (INIS)

    Zurron-Cifuentes, Oscar; Ortiz-Trujillo, Diego; Blanco-Fernandez, Luis A.

    2010-01-01

    Over the last few years the Juzbado Plant has developed and implemented several analysis methodologies to cope with specific issues regarding safety management. This paper describes the three most outstanding of them, so as to say, the Integrated Safety Analysis (ISA) project, the adaptation of the MARSSIM methodology for characterization surveys of radioactive contamination spots, and the programme for the Systematic Review of the Operational Conditions of the Safety Systems (SROCSS). Several reasons motivated the decision to implement such methodologies, such as Regulator requirements, operational experience and of course, the strong commitment of ENUSA to maintain the highest standards of nuclear industry on all the safety relevant activities. In this context, since 2004 ENUSA is undertaking the ISA project, which consists on a systematic examination of plant's processes, equipment, structures and personnel activities to ensure that all relevant hazards that could result in unacceptable consequences have been adequately evaluated and the appropriate protective measures have been identified. On the other hand and within the framework of a current programme to ensure the absence of radioactive contamination spots on unintended areas, the MARSSIM methodology is being applied as a tool to conduct the radiation surveys and investigation of potentially contaminated areas. Finally, the SROCSS programme was initiated earlier this year 2009 to assess the actual operating conditions of all the systems with safety relevance, aiming to identify either potential non-conformities or areas for improvement in order to ensure their high performance after years of operation. The following paragraphs describe the key points related to these three methodologies as well as an outline of the results obtained so far. (authors)

  3. Comparative analysis of safety related site characteristics

    International Nuclear Information System (INIS)

    Andersson, Johan

    2010-12-01

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

  4. Comparative analysis of safety related site characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Johan (ed.)

    2010-12-15

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

  5. Qualitative analysis in reliability and safety studies

    International Nuclear Information System (INIS)

    Worrell, R.B.; Burdick, G.R.

    1976-01-01

    The qualitative evaluation of system logic models is described as it pertains to assessing the reliability and safety characteristics of nuclear systems. Qualitative analysis of system logic models, i.e., models couched in an event (Boolean) algebra, is defined, and the advantages inherent in qualitative analysis are explained. Certain qualitative procedures that were developed as a part of fault-tree analysis are presented for illustration. Five fault-tree analysis computer-programs that contain a qualitative procedure for determining minimal cut sets are surveyed. For each program the minimal cut-set algorithm and limitations on its use are described. The recently developed common-cause analysis for studying the effect of common-causes of failure on system behavior is explained. This qualitative procedure does not require altering the fault tree, but does use minimal cut sets from the fault tree as part of its input. The method is applied using two different computer programs. 25 refs

  6. Applications of probabilistic risk analysis in nuclear criticality safety design

    International Nuclear Information System (INIS)

    Chang, J.K.

    1992-01-01

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

  7. Analysis of decision alternatives of the deep borehole filter restoration problem

    International Nuclear Information System (INIS)

    Abdildin, Yerkin G.; Abbas, Ali E.

    2016-01-01

    The energy problem is one of the biggest challenges facing the world in the 21st century. The nuclear energy is the fastest-growing contributor to the world energy and uranium mining is the primary step in its chain. One of the fundamental problems in the uranium extraction industry is the deep borehole filter restoration problem. This decision problem is very complex due to multiple objectives and various uncertainties. Besides the improvement of uranium production, the decision makers often need to meet internationally recognized standards (ISO 14001) of labor protection, safety measures, and preservation of environment. The problem can be simplified by constructing the multiattribute utility function, but the choice of the appropriate functional form requires the practical evaluation of different methods. In present work, we evaluate the alternatives of this complex problem by two distinct approaches for analyzing decision problems. The decision maker and the assessor is a Deputy Director General of a transnational corporation. - Highlights: • Analyzes 5 borehole recovery methods across the 4 most important attributes (criteria). • Considers financial, technological, environmental, and safety factors. • Compares two decision analysis approaches and the profit analysis. • Illustrates the assessments of the decision maker's preferences. • Determines that the assumption of independence of attributes yields imprecise recommendations.

  8. Capsule safety analysis of PRTF irradiation facility

    International Nuclear Information System (INIS)

    Suwarto

    2013-01-01

    Power Ramp Test Facility (PRTF) is an irradiation facility used for fuel testing of power reactor. PRTF has a capsule which is a test fuel rod container. During operation, pressurized water of 160 bars flows through in the capsule. Due to the high pressure it should be analyzed the impact of the capsule on reactor core safety. This analysis has purpose to calculate the ability of capsule pressure capacity. The analysis was carried out by calculating pressure capacity. From the calculating results it can be concluded that the capsule with pressure capacity of 438 bars will be safe to prevent the operation pressure of PRTF. (author)

  9. Ensuring Adequate Health and Safety Information for Decision Makers during Large-Scale Chemical Releases

    Science.gov (United States)

    Petropoulos, Z.; Clavin, C.; Zuckerman, B.

    2015-12-01

    The 2014 4-Methylcyclohexanemethanol (MCHM) spill in the Elk River of West Virginia highlighted existing gaps in emergency planning for, and response to, large-scale chemical releases in the United States. The Emergency Planning and Community Right-to-Know Act requires that facilities with hazardous substances provide Material Safety Data Sheets (MSDSs), which contain health and safety information on the hazardous substances. The MSDS produced by Eastman Chemical Company, the manufacturer of MCHM, listed "no data available" for various human toxicity subcategories, such as reproductive toxicity and carcinogenicity. As a result of incomplete toxicity data, the public and media received conflicting messages on the safety of the contaminated water from government officials, industry, and the public health community. Two days after the governor lifted the ban on water use, the health department partially retracted the ban by warning pregnant women to continue avoiding the contaminated water, which the Centers for Disease Control and Prevention deemed safe three weeks later. The response in West Virginia represents a failure in risk communication and calls to question if government officials have sufficient information to support evidence-based decisions during future incidents. Research capabilities, like the National Science Foundation RAPID funding, can provide a solution to some of the data gaps, such as information on environmental fate in the case of the MCHM spill. In order to inform policy discussions on this issue, a methodology for assessing the outcomes of RAPID and similar National Institutes of Health grants in the context of emergency response is employed to examine the efficacy of research-based capabilities in enhancing public health decision making capacity. The results of this assessment highlight potential roles rapid scientific research can fill in ensuring adequate health and safety data is readily available for decision makers during large

  10. Fuzzy multi-objective decision making on a low and intermediate level waste repository safety assessment

    International Nuclear Information System (INIS)

    Lemos, Francisco Luiz de; Deshpande, Ashok; Guimaraes, Lamartine

    2002-01-01

    Low and intermediate waste disposal facilities safety assessment is comprised of several steps from site selection , construction and operation to post-closure performance assessment. This is a multidisciplinary and complex task , and can not be analyzed by one expert only. This high complexity can lead to ambiguity and vagueness in information and consequently in the decision making process. In order to make the decision process clear and objective, there is the need to provide the decision makers with a clear and comprehensive picture of the whole process and, at the same time, simple and easily understandable by the public. This paper suggests the development of an inference system based on fuzzy decision making methodology. Fuzzy logic tools are specially suited to deal with ambiguous data by using language expressions. This process would be capable of integrating knowledge from various fields of environmental sciences. It has an advantage of keeping record of reasoning for each intermediate decision that lead to the final results which makes it more dependable and defensible as well. (author)

  11. Fuel reprocessing: safety analysis of extraction cycles

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. Decision analysis for dynamic accounting of nuclear material

    International Nuclear Information System (INIS)

    Shipley, J.P.

    1978-01-01

    Effective materials accounting for special nuclear material in modern fuel cycle facilities will depend heavily on sophisticated data analysis techniques. Decision analysis, which combines elements of estimation theory, decision theory, and systems analysis, is a framework well suited to the development and application of these techniques. Augmented by pattern-recognition tools such as the alarm-sequence chart, decision analysis can be used to reduce errors caused by subjective data evaluation and to condense large collections of data to a smaller set of more descriptive statistics. Application to data from a model plutonium nitrate-to-oxide conversion process illustrates the concepts

  13. System analysis of vehicle active safety problem

    Science.gov (United States)

    Buznikov, S. E.

    2018-02-01

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

  14. Multi-criteria analysis for evaluating the radiological and ecological safety measures in radioactive waste management

    International Nuclear Information System (INIS)

    Sazykina, T.G.; Kryshev, I.I.

    2006-01-01

    A methodological approach is presented for multicriterial evaluating the effectiveness of radiation ecological safety measures during radioactive waste management. The approach is based on multicriterial analysis with consideration of radiological, ecological, social, economical consequences of various safety measures. The application of the multicriterial approach is demonstrated taking as an example of decision-making on the most effective actions for rehabilitation of a water subject, contaminated with radionuclides [ru

  15. Multivariate Analysis of Household Decision Making, Contraceptive ...

    African Journals Online (AJOL)

    Nneka Umera-Okeke

    contraceptives and fertility behaviour of ever-married men in Nigeria. ... exposure. The study concluded that women empowerment in decision ... through the prevention of unwanted and unplanned births is one of the most effective .... visitors who slept in the household the previous night before the survey) were eligible ...

  16. A regret theory approach to decision curve analysis: a novel method for eliciting decision makers' preferences and decision-making.

    Science.gov (United States)

    Tsalatsanis, Athanasios; Hozo, Iztok; Vickers, Andrew; Djulbegovic, Benjamin

    2010-09-16

    Decision curve analysis (DCA) has been proposed as an alternative method for evaluation of diagnostic tests, prediction models, and molecular markers. However, DCA is based on expected utility theory, which has been routinely violated by decision makers. Decision-making is governed by intuition (system 1), and analytical, deliberative process (system 2), thus, rational decision-making should reflect both formal principles of rationality and intuition about good decisions. We use the cognitive emotion of regret to serve as a link between systems 1 and 2 and to reformulate DCA. First, we analysed a classic decision tree describing three decision alternatives: treat, do not treat, and treat or no treat based on a predictive model. We then computed the expected regret for each of these alternatives as the difference between the utility of the action taken and the utility of the action that, in retrospect, should have been taken. For any pair of strategies, we measure the difference in net expected regret. Finally, we employ the concept of acceptable regret to identify the circumstances under which a potentially wrong strategy is tolerable to a decision-maker. We developed a novel dual visual analog scale to describe the relationship between regret associated with "omissions" (e.g. failure to treat) vs. "commissions" (e.g. treating unnecessary) and decision maker's preferences as expressed in terms of threshold probability. We then proved that the Net Expected Regret Difference, first presented in this paper, is equivalent to net benefits as described in the original DCA. Based on the concept of acceptable regret we identified the circumstances under which a decision maker tolerates a potentially wrong decision and expressed it in terms of probability of disease. We present a novel method for eliciting decision maker's preferences and an alternative derivation of DCA based on regret theory. Our approach may be intuitively more appealing to a decision-maker, particularly

  17. A regret theory approach to decision curve analysis: A novel method for eliciting decision makers' preferences and decision-making

    Directory of Open Access Journals (Sweden)

    Vickers Andrew

    2010-09-01

    Full Text Available Abstract Background Decision curve analysis (DCA has been proposed as an alternative method for evaluation of diagnostic tests, prediction models, and molecular markers. However, DCA is based on expected utility theory, which has been routinely violated by decision makers. Decision-making is governed by intuition (system 1, and analytical, deliberative process (system 2, thus, rational decision-making should reflect both formal principles of rationality and intuition about good decisions. We use the cognitive emotion of regret to serve as a link between systems 1 and 2 and to reformulate DCA. Methods First, we analysed a classic decision tree describing three decision alternatives: treat, do not treat, and treat or no treat based on a predictive model. We then computed the expected regret for each of these alternatives as the difference between the utility of the action taken and the utility of the action that, in retrospect, should have been taken. For any pair of strategies, we measure the difference in net expected regret. Finally, we employ the concept of acceptable regret to identify the circumstances under which a potentially wrong strategy is tolerable to a decision-maker. Results We developed a novel dual visual analog scale to describe the relationship between regret associated with "omissions" (e.g. failure to treat vs. "commissions" (e.g. treating unnecessary and decision maker's preferences as expressed in terms of threshold probability. We then proved that the Net Expected Regret Difference, first presented in this paper, is equivalent to net benefits as described in the original DCA. Based on the concept of acceptable regret we identified the circumstances under which a decision maker tolerates a potentially wrong decision and expressed it in terms of probability of disease. Conclusions We present a novel method for eliciting decision maker's preferences and an alternative derivation of DCA based on regret theory. Our approach may

  18. The impact of safety and quality of health care on Chinese nursing career decision-making.

    Science.gov (United States)

    Zhu, Junhong; Rodgers, Sheila; Melia, Kath M

    2014-05-01

    The aim of the study was to understand why nurses leave nursing practice in China by exploring the process from recruitment to final exit. This report examines the impact of safety and quality of health care on nursing career decision-making from the leavers' perspective. The nursing shortage in China is more serious than in most developed countries, but the loss of nurses through voluntarily leaving nursing practice has not attracted much attention. This qualitative study draws on a grounded theory approach. In-depth interviews with 19 nurses who have left nursing practice and were theoretically sampled from one provincial capital city in Mainland China. 'Loss of confidence in the safety and quality of health care' became one of the main categories from all leavers' accounts of their decision to leave nursing practice. It emerged from three themes 'Perceiving risk in clinical practice', 'Recognising organisational barriers to safety' and 'Failing to meet expectations of patients'. The findings indicate that the essential work value of nursing to the leavers is the safety and quality of care for their patients. When nurses perceived that they could not fulfil this essential work value in their nursing practice, some of them could not accept the compromise to their value of nursing and left voluntarily to get away from the physical and mental stress. However, some nurses had to stay and accept the limitations on the safety and quality of health care. The study suggests that well-qualified nurses voluntarily leaving nursing practice is a danger signal for patients and hospitals, and has caused deterioration in nursing morale for both current and potential nursing workforces. It suggests that safety and quality of health care could be improved when individual nurses are empowered to exercise nursing autonomy with organisational and managerial support. The priority retention strategies need to remove organisational barriers to the safety and quality of health care

  19. Decision analysis interviews on protective actions in Finland supported by the RODOS system

    Energy Technology Data Exchange (ETDEWEB)

    Haemaelaeinen, R.P.; Lindstedt, M. [Helsinki Univ. of Technology, Otaniemi (Finland). System Analysis Lab.; Sinkko, K.; Ammann, M. [Radiation and Nuclear Safety Authority, Helsinki (Finland); Salo, A. [Lepolantie 54, Helsinki (Finland)

    2000-03-01

    This work was undertaken in order to study the utilisation of decision analysis interviews and of the RODOS system when planning protective actions in the case of a nuclear accident. Six decision analysis interview meetings were organised. Interviewees were competent national safety authorities and technical level decision-makers, i.e., those who are responsible for drawing up advice or making presentations of matters to decision-makers responsible for the practical implementation of the actions. The theme of the meetings was to study how uncertainties could be included in the decision-making process and whether pre-structured generic attributes and value trees would help this process and save time. The approach was to present a generic value tree, a decision table and a selected information package at the beginning of the interviews. The interviewees then examined the suggested value tree in order to ensure that no important factors have been omitted and they made changes when necessary. Also, the decision table was examined and altered by some participants and some of them asked for further information on some issues. But all in all the selected approach allowed for more time and effort to be allocated to value trade-offs and elicitation of risk attitudes. All information was calculated with the support of the RODOS system. Predefined value trees were found to ensure that all relevant factors are considered. The participants also felt that RODOS could provide the required information but, as in previous RODOS exercises, they found it more problematic to use decision analysis methods when planning countermeasures in the early phase of a nuclear accident. Furthermore, it was again noted that understanding the actual meaning 'soft' attributes, such as socio-psychological impacts, was not a straightforward issue. Consequently, the definition of attributes and training in advance would be beneficial. The incorporation of uncertainties also proved to be

  20. Decision analysis interviews on protective actions in Finland supported by the RODOS system

    International Nuclear Information System (INIS)

    Haemaelaeinen, R.P.; Lindstedt, M.; Salo, A.

    2000-03-01

    This work was undertaken in order to study the utilisation of decision analysis interviews and of the RODOS system when planning protective actions in the case of a nuclear accident. Six decision analysis interview meetings were organised. Interviewees were competent national safety authorities and technical level decision-makers, i.e., those who are responsible for drawing up advice or making presentations of matters to decision-makers responsible for the practical implementation of the actions. The theme of the meetings was to study how uncertainties could be included in the decision-making process and whether pre-structured generic attributes and value trees would help this process and save time. The approach was to present a generic value tree, a decision table and a selected information package at the beginning of the interviews. The interviewees then examined the suggested value tree in order to ensure that no important factors have been omitted and they made changes when necessary. Also, the decision table was examined and altered by some participants and some of them asked for further information on some issues. But all in all the selected approach allowed for more time and effort to be allocated to value trade-offs and elicitation of risk attitudes. All information was calculated with the support of the RODOS system. Predefined value trees were found to ensure that all relevant factors are considered. The participants also felt that RODOS could provide the required information but, as in previous RODOS exercises, they found it more problematic to use decision analysis methods when planning countermeasures in the early phase of a nuclear accident. Furthermore, it was again noted that understanding the actual meaning 'soft' attributes, such as socio-psychological impacts, was not a straightforward issue. Consequently, the definition of attributes and training in advance would be beneficial. The incorporation of uncertainties also proved to be difficult and

  1. Decision making analysis of walnut seedling production on a small ...

    African Journals Online (AJOL)

    The decision has to be made between those three alternatives aiming at achievement of optimal/best economic result for the family farm. Summarizing results obtained from the decision tree, simulation and sensitivity analysis, the optimal solution for the family farm should be to continue production of walnut seedlings with ...

  2. Household consumption, female employment and fertility decisions; A microeconometric analysis

    NARCIS (Netherlands)

    Kalwij, A.S.

    1999-01-01

    This thesis is mainly concerned with a simultaneous analysis of the economic determinants of female employment and fertility decisions on a household level in the Netherlands. In particular, this thesis is interested in the role of the employment decisions of women in the observed behavior that

  3. ARIES-AT safety design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D.A. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States)]. E-mail: David.Petti@inl.gov; Merrill, B.J. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Moore, R.L. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Longhurst, G.R. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); El-Guebaly, L. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Mogahed, E. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Henderson, D. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Wilson, P. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Abdou, A. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2006-01-15

    ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement such as a loss of vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement under average weather conditions. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, both the volume of the component and its hazard are used to classify the waste. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.

  4. Decision Analysis of Dynamic Spectrum Access Rules

    Energy Technology Data Exchange (ETDEWEB)

    Juan D. Deaton; Luiz A. DaSilva; Christian Wernz

    2011-12-01

    A current trend in spectrum regulation is to incorporate spectrum sharing through the design of spectrum access rules that support Dynamic Spectrum Access (DSA). This paper develops a decision-theoretic framework for regulators to assess the impacts of different decision rules on both primary and secondary operators. We analyze access rules based on sensing and exclusion areas, which in practice can be enforced through geolocation databases. Our results show that receiver-only sensing provides insufficient protection for primary and co-existing secondary users and overall low social welfare. On the other hand, using sensing information between the transmitter and receiver of a communication link, provides dramatic increases in system performance. The performance of using these link end points is relatively close to that of using many cooperative sensing nodes associated to the same access point and large link exclusion areas. These results are useful to regulators and network developers in understanding in developing rules for future DSA regulation.

  5. PATIENT-CENTERED DECISION MAKING: LESSONS FROM MULTI-CRITERIA DECISION ANALYSIS FOR QUANTIFYING PATIENT PREFERENCES.

    Science.gov (United States)

    Marsh, Kevin; Caro, J Jaime; Zaiser, Erica; Heywood, James; Hamed, Alaa

    2018-01-01

    Patient preferences should be a central consideration in healthcare decision making. However, stories of patients challenging regulatory and reimbursement decisions has led to questions on whether patient voices are being considered sufficiently during those decision making processes. This has led some to argue that it is necessary to quantify patient preferences before they can be adequately considered. This study considers the lessons from the use of multi-criteria decision analysis (MCDA) for efforts to quantify patient preferences. It defines MCDA and summarizes the benefits it can provide to decision makers, identifies examples of MCDAs that have involved patients, and summarizes good practice guidelines as they relate to quantifying patient preferences. The guidance developed to support the use of MCDA in healthcare provide some useful considerations for the quantification of patient preferences, namely that researchers should give appropriate consideration to: the heterogeneity of patient preferences, and its relevance to decision makers; the cognitive challenges posed by different elicitation methods; and validity of the results they produce. Furthermore, it is important to consider how the relevance of these considerations varies with the decision being supported. The MCDA literature holds important lessons for how patient preferences should be quantified to support healthcare decision making.

  6. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Lee, Y. B.; Kwon, Y. M.; Suk, S. D.

    2005-03-01

    The MATRA-LMR-FB has been developed internally for the damage prevention as well as the safety assessment during a channel blockage accident and, as a the result, the quality of the code becomes comparable to that developed in the leading countries. For a code-to-code comparison, KAERI could have access to the SASSYS-1 through a bilateral collaboration between KAERI and ANL. The study could bring into the reliability improvements both on the reactivity models in the SSC-K and on the SSC-K prediction capability. It finally leads to the completion of the SSC-K version 1.3 resulting from the qualitative and quantitative code-to-code comparison. The preliminary analysis for a metal fueled LMR could also become possible with the MELT-III and the VENUS-II, which had originally been developed for the HCDA analysis with an oxidized fuel, by developing the relevant models For the development of the safety evaluation technology, the safety limits have been set up, and the analyses of the internal and external channel blockages in an assembly have also been performed. Besides, the more reliable analysis results on the key design concepts could be obtained by way of the methodology improvement resulting from the qualitative and quantitative comparison study. For an efficient and systematic control of the main project, the integration of the developed technologies and the establishment of their data base have been pursued. It has gone through the development of the process control with taking account of interfaces among the sub-projects, the overall coordination of the developed technologies, the data base for the design products, and so on

  7. Incorporation of advanced accident analysis methodology into safety analysis reports

    International Nuclear Information System (INIS)

    2003-05-01

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

  8. ACRR fuel storage racks criticality safety analysis

    International Nuclear Information System (INIS)

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

    1997-10-01

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

  9. Safety analysis reports. Current status (third key report)

    International Nuclear Information System (INIS)

    1999-01-01

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

  10. Ignalina Safety Analysis Group's report for the year 1998

    International Nuclear Information System (INIS)

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

    1999-02-01

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

  11. ESSAA: Embedded system safety analysis assistant

    Science.gov (United States)

    Wallace, Peter; Holzer, Joseph; Guarro, Sergio; Hyatt, Larry

    1987-01-01

    The Embedded System Safety Analysis Assistant (ESSAA) is a knowledge-based tool that can assist in identifying disaster scenarios. Imbedded software issues hazardous control commands to the surrounding hardware. ESSAA is intended to work from outputs to inputs, as a complement to simulation and verification methods. Rather than treating the software in isolation, it examines the context in which the software is to be deployed. Given a specified disasterous outcome, ESSAA works from a qualitative, abstract model of the complete system to infer sets of environmental conditions and/or failures that could cause a disasterous outcome. The scenarios can then be examined in depth for plausibility using existing techniques.

  12. Rankine bottoming cycle safety analysis. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lewandowski, G.A.

    1980-02-01

    Vector Engineering Inc. conducted a safety and hazards analysis of three Rankine Bottoming Cycle Systems in public utility applications: a Thermo Electron system using Fluorinal-85 (a mixture of 85 mole % trifluoroethanol and 15 mole % water) as the working fluid; a Sundstrand system using toluene as the working fluid; and a Mechanical Technology system using steam and Freon-II as the working fluids. The properties of the working fluids considered are flammability, toxicity, and degradation, and the risks to both plant workers and the community at large are analyzed.

  13. Mechanistic facility safety and source term analysis

    International Nuclear Information System (INIS)

    PLYS, M.G.

    1999-01-01

    A PC-based computer program was created for facility safety and source term analysis at Hanford The program has been successfully applied to mechanistic prediction of source terms from chemical reactions in underground storage tanks, hydrogen combustion in double contained receiver tanks, and proccss evaluation including the potential for runaway reactions in spent nuclear fuel processing. Model features include user-defined facility room, flow path geometry, and heat conductors, user-defined non-ideal vapor and aerosol species, pressure- and density-driven gas flows, aerosol transport and deposition, and structure to accommodate facility-specific source terms. Example applications are presented here

  14. 324 building safety analysis report supplement

    International Nuclear Information System (INIS)

    Dodd, A.O.; Wittenbrock, N.G.

    1977-01-01

    Process engineering designs, major equipment and plant facilities to be utilized in commercial nuclear waste preparation and vitrification in the 324 Radiochemical Engineering Building are reviewed with regard to accident potential and consequences. This Safety Analysis Report Supplement compares calculated environmental doses anticipated from the Commercial Nuclear Waste Vitrification Project (CNWVP) routine operations with the average doses from past waste management operations conducted at the Hanford Project and finds them to be significantly less. The calculated CNWVP environmental doses are found to be far below presently applicable ERDA standards and standards proposed by the EPA for nuclear power operations

  15. Preliminary safety analysis of the Gorleben site

    International Nuclear Information System (INIS)

    Bracke, G.; Fischer-Appelt, K.

    2014-01-01

    The safety requirements governing the final disposal of heat-generating radioactive waste in Germany were implemented by the Federal Ministry of Environment, Natural Conservation and Nuclear Safety (BMU) in 2010. The Ministry considers as a fundamental objective the protection of man and the environment against the hazards of radioactive waste. Unreasonable burdens and obligation for future generations shall be avoided. The main safety principles are concentration and inclusion of radioactive and other pollutants in a containment-providing rock zone. Any release of radioactive nuclides may increase the risk for men and the environment only negligibly compared to natural radiation exposure. No intervention or maintenance work shall be necessary in the post-closure phase. Retrieval/recovery of the waste shall be possible up to 500 years after closure. The Gorleben salt dome has been discussed since the 1970's as a possible repository site for heat-generating radioactive waste in Germany. The objective of the project preliminary safety analysis of the Gorleben site (VSG) was to assess if repository concepts at the Gorleben site or other sites with a comparable geology could comply with these requirements based on currently available knowledge (Fischer-Appelt, 2013; Bracke, 2013). In addition to this it was assessed if methodological approaches can be used for a future site selection procedure and which technological and conceptual considerations can be transferred to other geological situations. The objective included the compilation and review of the available exploration data of the Gorleben site and on disposal in salt rock, the development of repository designs, and the identification of the needs for future R and D work and further site investigations. (authors)

  16. A method for analysis of nuclear power plant operators' decision making in simulated disturbance situations

    International Nuclear Information System (INIS)

    1995-01-01

    An analysis method has been developed for analysis of nuclear power plant operators' decision making in simulated disturbance situations. The aim of the analysis is to investigate operators' orientation which is expected to manifest itself as collective strategies in utilization of resources of decision making. Resources analyzed here are different information sources and, in addition, collaborative resources like communication and participation. The cognitive approach on the basis of the method considers decision making as collective construction of common interpretation of available information. Utilization of information is evaluated with respect to operative context. This is made with help of conceptualization of the disturbance situation from the decision making point of view and by construction of operative reference for activity. The latter means conceptualization of the situation from the safety point of view and also consideration of other boundary constraints of decision making, i.e. economical and technical aspects. The analysis method is intended to be used in routine simulator training in nuclear power plants. By virtue of its contextual and dynamical approach it makes the developing nature of activity visible. Cumulation and distribution of knowledge of decision making as developing activity, controlled by orientation and boundary constraints of process control, is expected to improve operational culture of a plant organization. (author). 2 refs, 1 fig

  17. Safety- and risk analysis activities in other areas than the nuclear industry

    International Nuclear Information System (INIS)

    Kozine, I.; Duijm, N.J.; Lauridsen, K.

    2000-12-01

    The report gives an overview of the legislation within the European Union in the field of major industrial hazards and gives examples of decision criteria applied in a number of European countries when judging the acceptability of an activity. Furthermore, the report mentions a few methods used in the analysis of the safety of chemical installations. (au)

  18. Role of seismic PRA in seismic safety decisions of nuclear power plants

    International Nuclear Information System (INIS)

    Ravindra, M.K.; Kennedy, R.P.; Sues, R.H.

    1985-01-01

    This paper highlights the important roles that seismic probabilistic risk assessments (PRAs) can play in the seismic safety decisions of nuclear power plants. If a seismic PRA has been performed for a plant, its results can be utilized to evaluate the seismic capability beyond the safe shutdown event (SSE). Seismic fragilities of key structures and equipment, fragilities of dominant plant damage states and the frequencies of occurrence of these plant damage states are reviewed to establish the seismic safety of the plant beyond the SSE level. Guidelines for seismic margin reviews and upgrading may be developed by first identifying the generic classes of structures and equipment that have been shown to be dominant risk contributors in the completed seismic PRAs, studying the underlying causes for their contribution and examining why certain other items (e.g., piping) have not proved to be high-risk-contributors

  19. Decision analysis for the siting of nuclear power plants: the relevance of multiattribute utility theory

    International Nuclear Information System (INIS)

    Keeney, R.L; Nair, K.

    1975-01-01

    The necessity for improved decision making concerning the siting and licensing of major power facilities has been accelerated in the past decade by the increased environmental consciousness of the public and by the energy crisis. These problems are exceedingly complex due to their multiple objective nature, the many interest groups, the long-range time horizons, and the inherent uncertainties of the potential impacts of any decision. Along with the relatively objective economic and engineering concerns, clearly the more subjective factors involving safety, environmental, and social issues are crucial to the problem. Hence, the professional judgments and knowledge of experts in these areas should be utilized in analyses of siting decisions. Likewise, the preferences of the general public, as consumers, the utility companies, as builders and operators of power plant facilities, and environmentalists and the government must be accounted for in analyzing power plant siting and licensing issues. We advocate an approach for formally articulating the experts' judgments and the decision makers' preferences, both of which are clearly subjective, and processing these along with the more objective considerations in a logical manner to acquire the implications for decision making. The appropriateness and application of decision analysis for power plant location decisions is discussed and illustrated. Emphasis is placed on the assessment of the decision maker's preferences and tradeoffs concerning multiple objectives. (U.S.)

  20. Decision analysis of shoreline protection under climate change uncertainty

    Science.gov (United States)

    Chao, Philip T.; Hobbs, Benjamin F.

    1997-04-01

    If global warming occurs, it could significantly affect water resource distribution and availability. Yet it is unclear whether the prospect of such change is relevant to water resources management decisions being made today. We model a shoreline protection decision problem with a stochastic dynamic program (SDP) to determine whether consideration of the possibility of climate change would alter the decision. Three questions are addressed with the SDP: (l) How important is climate change compared to other uncertainties?, (2) What is the economic loss if climate change uncertainty is ignored?, and (3) How does belief in climate change affect the timing of the decision? In the case study, sensitivity analysis shows that uncertainty in real discount rates has a stronger effect upon the decision than belief in climate change. Nevertheless, a strong belief in climate change makes the shoreline protection project less attractive and often alters the decision to build it.

  1. The Decision of Information Safety Problems at Processing of the Biometric Personal Data

    Directory of Open Access Journals (Sweden)

    Y. G. Gorshkov

    2010-03-01

    Full Text Available The requirements imposed on transfer by the personal biometric information in systems and communication networks according to Federal Law № 152 “Personal data” are defined. Lacks of used decisions protection of such biometric data, as the test speech information, including parameters of a speech path, and also acoustic signals of tones and noise of heart of the person on an example of telemedicine systems construction with the using of a network telephone channels general using and wireless networks Wi-Fi are considered. Directions of works are formulated on safety of the personal biometric data transferred in telecommunication systems.

  2. Improving the Efficiency of Administrative Decision-Making when Monitoring Reliability and Safety of Oil and Gas Equipment

    Directory of Open Access Journals (Sweden)

    Zemenkova Maria

    2016-01-01

    Full Text Available Methodology of rapid assessment of reliability index was developed based on system analysis of technological parameters. Within functioning of on-line monitoring system of reliability index of industrial facility this method allows to increase efficiency of making managerial decisions on technical and preventive maintenance. The technique is based on the analysis of technological parameters of operational modes of pipeline transport facilities registered by dispatcher controls. The created technique can be used by the operating, research, design institutes and oil and gas transport enterprises when declaring industrial safety. The received mathematical models allow federal services of supervision, the independent expert organizations to predict the development of reliability in the registered block of dispatching data either in real time mode, or taking into account the dynamics of service conditions of the object.

  3. Safety strategy and safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1976-01-01

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

  4. A safety-critical decision support system evaluation using situation awareness and workload measures

    International Nuclear Information System (INIS)

    Naderpour, Mohsen; Lu, Jie; Zhang, Guangquan

    2016-01-01

    To ensure the safety of operations in safety-critical systems, it is necessary to maintain operators' situation awareness (SA) at a high level. A situation awareness support system (SASS) has therefore been developed to handle uncertain situations [1]. This paper aims to systematically evaluate the enhancement of SA in SASS by applying a multi-perspective approach. The approach consists of two SA metrics, SAGAT and SART, and one workload metric, NASA-TLX. The first two metrics are used for the direct objective and subjective measurement of SA, while the third is used to estimate operator workload. The approach is applied in a safety-critical environment called residue treater, located at a chemical plant in which a poor human-system interface reduced the operator's SA and caused one of the worst accidents in US history. A counterbalanced within-subjects experiment is performed using a virtual environment interface with and without the support of SASS. The results indicate that SASS improves operators' SA, and specifically has benefits for SA levels 2 and 3. In addition, it is concluded that SASS reduces operator workload, although further investigations in different environments with a larger number of participants have been suggested. - Highlights: • The suitability of a cognitive decision support system is investigated. • An evaluation approach considering situation awareness and workload measures is proposed. • A computerized system based on the proposed approach is implemented. • The implemented system is used in a safety-critical environment.

  5. Probabilistic safety analysis applied to RBMK reactors

    International Nuclear Information System (INIS)

    Gerez Martin, L.; Fernandez Ramos, P.

    1995-01-01

    The project financed by the European Union ''Revision of RBMK Reactor Safety was divided into nine Topic Groups dealing with different aspects of safety. The area covered by Topic Group 9 was Probabilistic Safety Analysis. TG9 will have touched on some of the problems discussed by other groups, although in terms of the systematic quantification of the impact of design characteristics and RBMK reactor operating practices on the risk of core damage. On account of the reduced time scale and the resources available for the project, the analysis was made using a simplified method based on the results of PSAs conducted in Western countries and on the judgement of the group members. The simplifies method is based on the concepts of Qualification, Redundancy and Automatic Actuation of the systems considered. PSA experience shows that systems complying with the above-mentioned concepts have a failure probability of 1.0E-3 when redundancy is simple, ie two similar equipment items capable of carrying out the same function. In general terms, this value can be considered to be dominated by potential common cause failures. The value considered above changes according to factors that have a positive effect upon it, such as an additional redundancy with a different equipment item (eg a turbo pumps and a motor pump), individual trains with good separations, etc, or a negative effect, such as the absence of suitable periodical tests, the need for operators to perform manual operations, etc. Similarly, possible actions required by the operator during accident sequences are assigned failure probability values between 1 and 1.0E-4, according to the complexity of the action (including local actions to be performed outside the control room) and the time available

  6. Uncertainty about probability: a decision analysis perspective

    International Nuclear Information System (INIS)

    Howard, R.A.

    1988-01-01

    The issue of how to think about uncertainty about probability is framed and analyzed from the viewpoint of a decision analyst. The failure of nuclear power plants is used as an example. The key idea is to think of probability as describing a state of information on an uncertain event, and to pose the issue of uncertainty in this quantity as uncertainty about a number that would be definitive: it has the property that you would assign it as the probability if you knew it. Logical consistency requires that the probability to assign to a single occurrence in the absence of further information be the mean of the distribution of this definitive number, not the medium as is sometimes suggested. Any decision that must be made without the benefit of further information must also be made using the mean of the definitive number's distribution. With this formulation, they find further that the probability of r occurrences in n exchangeable trials will depend on the first n moments of the definitive number's distribution. In making decisions, the expected value of clairvoyance on the occurrence of the event must be at least as great as that on the definitive number. If one of the events in question occurs, then the increase in probability of another such event is readily computed. This means, in terms of coin tossing, that unless one is absolutely sure of the fairness of a coin, seeing a head must increase the probability of heads, in distinction to usual thought. A numerical example for nuclear power shows that the failure of one plant of a group with a low probability of failure can significantly increase the probability that must be assigned to failure of a second plant in the group

  7. Combining morphological analysis and Bayesian Networks for strategic decision support

    CSIR Research Space (South Africa)

    De Waal, AJ

    2007-12-01

    Full Text Available Morphological analysis (MA) and Bayesian networks (BN) are two closely related modelling methods, each of which has its advantages and disadvantages for strategic decision support modelling. MA is a method for defining, linking and evaluating...

  8. SPATIAL ANALYSIS AND DECISION ASSISTANCE (SADA) TRAINING COURSE

    Science.gov (United States)

    Spatial Analysis and Decision Assistance (SADA) is a Windows freeware program that incorporates tools from environmental assessment into an effective problem-solving environment. SADA was developed by the Institute for Environmental Modeling at the University of Tennessee and inc...

  9. A Bayesian Network methodology for railway risk, safety and decision support

    OpenAIRE

    Mahboob, Qamar

    2014-01-01

    For railways, risk analysis is carried out to identify hazardous situations and their consequences. Until recently, classical methods such as Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) were applied in modelling the linear and logically deterministic aspects of railway risks, safety and reliability. However, it has been proven that modern railway systems are rather complex, involving multi-dependencies between system variables and uncertainties about these dependencies. For train ...

  10. 242-A evaporator safety analysis report

    International Nuclear Information System (INIS)

    CAMPBELL, T.A.

    1999-01-01

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR

  11. Short course on system safety analysis

    International Nuclear Information System (INIS)

    Sudmann, R.H.

    1992-01-01

    This course provides and introduction to methods generally used in safety analysis and accident investigation. It is a non-mathematical approach, directed toward a casual user. The participant will learn techniques allowing them to dissect a system or incident in order identify real or potential safety problems. These techniques will be applied to analyze events which have occurred within DOE facilities. As a manager or staff person with general oversight responsibilities, the participant should gain an awareness of the big picture and not just ''dig for facts.'' This can be accomplished by being alert and responsive to the atmosphere and condition of the plant; mood and impression of the worker and the behavioral climate. The techniques taught in the course can be used to identify critical areas or indicators. These indicators will signal problems before the ''facts'' will. Analysis techniques taught are used to gauge the breadth of the ''forest'' and not necessarily to identify the trees. For this course includes a technical background with experience in a chemical processing operations and a knowledge of basic chemistry and engineering is desirable. The course should help in a present or future assignment in an oversight role

  12. 242-A evaporator safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    CAMPBELL, T.A.

    1999-05-17

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR.

  13. Safety and safety analysis. From CP1 to Fukushima

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2012-01-01

    The safety of nuclear installations has been a serious concern starting from the days of infancy of this technology. When Fermi and co-workers built the first nuclear reactor in 1941, the Chicago Pile-1 or CP1 at the University of Chicago, some basic safety principles still in use today were already part of this very simple experiment. During the fast-growth period in the 1960ies, a number of NPP systems were conceived, tested and some of them built, mainly in the US and in the Soviet Union, but also in the UK, in France and in Canada, before just a handful of nuclear systems dominated: the LWRs conquered some 3 quarters of the world market and their dominance continues till today. The fission process has been amazingly well ''designed'' by nature: a remarkably simple to produce, self-sustained reaction that can be easily controlled, modulated and adjusted by a variety of available materials. Fission leads to large release of energy that can be easily collected and transformed into useful work. The process has only a major drawback, the inexorable production and accumulation in the core of the radioactive fission products that also produce decay heat. Criticality considerations put apart, the major goal of reactor safety is the confinement and cooling of these fission products. Although safety has been a major concern from the very first nuclear developments, feedback and actions following incidents and accidents have contributed to continuous enhancements. In particular, the three major nuclear accidents, TMI, Chernobyl and Fukushima had or will hopefully have in the future major impacts on safety improvements. Lessons learned from TMI have greatly enhanced the safety of LWRs, while Chernobyl triggered a number of radio-ecology studies and improved the readiness for radiological crisis management. It is hoped that Fukushima will be the trigger for much stronger international oversight and harmonization of safety practices, something that has already been launched

  14. Safety and safety analysis. From CP1 to Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, George [ASCOMP GmbH, Zurich (Switzerland)

    2012-02-15

    The safety of nuclear installations has been a serious concern starting from the days of infancy of this technology. When Fermi and co-workers built the first nuclear reactor in 1941, the Chicago Pile-1 or CP1 at the University of Chicago, some basic safety principles still in use today were already part of this very simple experiment. During the fast-growth period in the 1960ies, a number of NPP systems were conceived, tested and some of them built, mainly in the US and in the Soviet Union, but also in the UK, in France and in Canada, before just a handful of nuclear systems dominated: the LWRs conquered some 3 quarters of the world market and their dominance continues till today. The fission process has been amazingly well ''designed'' by nature: a remarkably simple to produce, self-sustained reaction that can be easily controlled, modulated and adjusted by a variety of available materials. Fission leads to large release of energy that can be easily collected and transformed into useful work. The process has only a major drawback, the inexorable production and accumulation in the core of the radioactive fission products that also produce decay heat. Criticality considerations put apart, the major goal of reactor safety is the confinement and cooling of these fission products. Although safety has been a major concern from the very first nuclear developments, feedback and actions following incidents and accidents have contributed to continuous enhancements. In particular, the three major nuclear accidents, TMI, Chernobyl and Fukushima had or will hopefully have in the future major impacts on safety improvements. Lessons learned from TMI have greatly enhanced the safety of LWRs, while Chernobyl triggered a number of radio-ecology studies and improved the readiness for radiological crisis management. It is hoped that Fukushima will be the trigger for much stronger international oversight and harmonization of safety practices, something that has

  15. The design and verification of probabilistic safety analysis platform NFRisk

    International Nuclear Information System (INIS)

    Hu Wenjun; Song Wei; Ren Lixia; Qian Hongtao

    2010-01-01

    To increase the technical ability in Probabilistic Safety Analysis (PSA) field in China,it is necessary and important to study and develop indigenous professional PSA platform. Following such principle as 'from structure simplification to modulization to production of cut sets to minimum of cut sets', the algorithms, including simplification algorithm, modulization algorithm, the algorithm of conversion from fault tree to binary decision diagram (BDD), the solving algorithm of cut sets, the minimum algorithm of cut sets, and so on, were designed and developed independently; the design of data management and operation platform was completed all alone; the verification and validation of NFRisk platform based on 3 typical fault trees was finished on our own. (authors)

  16. A decision analysis of an exploratory studies facility

    International Nuclear Information System (INIS)

    Merkhofer, M.W.; Gnirk, P.

    1991-01-01

    An Exploratory Studies Facility (ESF) is planned to support the characterization of a potential site for a high-level nuclear waste repository at Yucca Mountain, NV. The selection of a design for the ESF is a critical decision, because the ESF design may affect the accuracy of characterization testing and subsequent repository design. The assist the design process, a comparative evaluation was conducted to rank 34 alternative relied on techniques from formal decision analysis, including decision trees and multiattribute utility analysis (MUA). The results helped to identify favorable design features and convinced the Department of Energy to adopt the top-ranked option as the preferred ESF design

  17. DASS: A decision aid integrating the safety parameter display system and emergency functional recovery procedures. Final report

    International Nuclear Information System (INIS)

    Johnson, S.E.

    1984-08-01

    Using a stand-alone developmental test-bed consisting of a minicomputer and a high-resolution color graphics computer, displays and supporting software incorporating advanced on-line decision-aid concepts were developed and evaluated. The advanced concepts embodied in displays designed for the operating crew of a PWR plant include: (1) an integrated display format which supports a top-down approach to problem detection, recovery planning, and control; (2) introduction of nonobservable plant parameters derived from first principles mass and energy balances as part of the displayed information; and (3) systematic processing and display of key success path (plant safety system) attributes. The prototype system, referred to as the PWR-DASS (Disturbance Analysis and Surveillance System), consists of 18 displays targeted for principal use by the control room systems manager. PWR-DASS was conceived to fulfill an operational void not fully supported by safety parameter display systems or reformulated emergency procedure guidelines. The results from the evaluation by licensed operators suggest that organization and display of desired critical safety function and success path information as incorporated in the PWR-DASS prototype can support the systems manager's overview. The results also point to the need for several refinements required for a field grade system, and to the need for a simulator-based evaluation of the prototype or its successor. (author)

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  19. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    1997-01-01

    The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

  20. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  1. Endogenous Risks and Learning in Climate Change Decision Analysis

    International Nuclear Information System (INIS)

    O'Neill, B.C.; Ermoliev, Y.; Ermolieva, T.

    2005-01-01

    We analyze the effects of risks and learning on climate change decisions. A two-stage, dynamic, climate change stabilization problem is formulated. The explicit incorporation of ex-post learning induces risk aversion among ex-ante decisions, which is characterized in linear models by VaR- (Value at Risk) and CVaR-type risk (Conditional Value at Risk) measures. Combined with explicit introduction of 'safety' constraints, it creates a 'hit-or-miss' type decision making situation and shows that, even in linear models, learning may lead to either less or more restrictive ex-ante emission reductions. We analyze stylized elements of the model in order to identify the key factors driving outcomes, in particular, the critical role of quantiles of probability distributions characterizing key uncertainties

  2. Review of decision methodologies for evaluating regulatory actions affecting public health and safety. [Nuclear industry site selection

    Energy Technology Data Exchange (ETDEWEB)

    Hendrickson, P.L.; McDonald, C.L.; Schilling, A.H.

    1976-12-01

    This report examines several aspects of the problems and choices facing the governmental decision maker who must take regulatory actions with multiple decision objectives and attributes. Particular attention is given to the problems facing the U.S. Nuclear Regulatory Commission (NRC) and to the decision attribute of chief concern to NRC, the protection of human health and safety, with emphasis on nuclear power plants. The study was undertaken to provide background information for NRC to use in refining its process of value/impact assessment of proposed regulatory actions. The principal conclusion is that approaches to rationally consider the value and impact of proposed regulatory actions are available. These approaches can potentially improve the decision-making process and enable the agency to better explain and defend its decisions. They also permit consistent examination of the impacts, effects of uncertainty and sensitivity to various assumptions of the alternatives being considered. Finally, these approaches can help to assure that affected parties are heard and that technical information is used appropriately and to the extent possible. The principal aspects of the regulatory decision problem covered in the report are: the legal setting for regulatory decisions which affect human health and safety, elements of the decision-making process, conceptual approaches to decision making, current approaches to decision making in several Federal agencies, and the determination of acceptable risk levels.

  3. Decision Analysis for Metric Selection on a Clinical Quality Scorecard.

    Science.gov (United States)

    Guth, Rebecca M; Storey, Patricia E; Vitale, Michael; Markan-Aurora, Sumita; Gordon, Randolph; Prevost, Traci Q; Dunagan, Wm Claiborne; Woeltje, Keith F

    2016-09-01

    Clinical quality scorecards are used by health care institutions to monitor clinical performance and drive quality improvement. Because of the rapid proliferation of quality metrics in health care, BJC HealthCare found it increasingly difficult to select the most impactful scorecard metrics while still monitoring metrics for regulatory purposes. A 7-step measure selection process was implemented incorporating Kepner-Tregoe Decision Analysis, which is a systematic process that considers key criteria that must be satisfied in order to make the best decision. The decision analysis process evaluates what metrics will most appropriately fulfill these criteria, as well as identifies potential risks associated with a particular metric in order to identify threats to its implementation. Using this process, a list of 750 potential metrics was narrowed to 25 that were selected for scorecard inclusion. This decision analysis process created a more transparent, reproducible approach for selecting quality metrics for clinical quality scorecards. © The Author(s) 2015.

  4. METHODOLOGY FOR ANALYSIS OF DECISION MAKING IN AIR NAVIGATION SYSTEM

    Directory of Open Access Journals (Sweden)

    Volodymyr Kharchenko

    2011-03-01

    Full Text Available Abstract. In the research of Air Navigation System as a complex socio-technical system the methodologyof analysis of human-operator's decision-making has been developed. The significance of individualpsychologicalfactors as well as the impact of socio-psychological factors on the professional activities of ahuman-operator during the flight situation development from normal to catastrophic were analyzed. On thebasis of the reflexive theory of bipolar choice the expected risks of decision-making by the Air NavigationSystem's operator influenced by external environment, previous experience and intentions were identified.The methods for analysis of decision-making by the human-operator of Air Navigation System usingstochastic networks have been developed.Keywords: Air Navigation System, bipolar choice, human operator, decision-making, expected risk, individualpsychologicalfactors, methodology of analysis, reflexive model, socio-psychological factors, stochastic network.

  5. Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

    International Nuclear Information System (INIS)

    Hayakawa, Masahiko

    1981-01-01

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

  6. An integrated framework for cost- benefit analysis in road safety projects using AHP method

    Directory of Open Access Journals (Sweden)

    Mahsa Mohamadian

    2011-10-01

    Full Text Available Cost benefit analysis (CBA is a useful tool for investment decision-making from economic point of view. When the decision involves conflicting goals, the multi-attribute analysis approach is more capable; because there are some social and environmental criteria that cannot be valued or monetized by cost benefit analysis. The complex nature of decision-making in road safety normally makes it difficult to reach a single alternative solution that can satisfy all decision-making problems. Generally, the application of multi-attribute analysis in road sector is promising; however, the applications are in preliminary stage. Some multi-attribute analysis techniques, such as analytic hierarchy process (AHP have been widely used in practice. This paper presents an integrated framework with CBA and AHP methods to select proper alternative in road safety projects. The proposed model of this paper is implemented for a case study of improving a road to reduce the accidents in Iran. The framework is used as an aid to cost benefit tool in road safety projects.

  7. Safety analysis of the VLJ repository

    International Nuclear Information System (INIS)

    Vieno, T.; Nordman, H.

    1991-05-01

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

  8. PCA safety data review after clinical decision support and smart pump technology implementation.

    Science.gov (United States)

    Prewitt, Judy; Schneider, Susan; Horvath, Monica; Hammond, Julia; Jackson, Jason; Ginsberg, Brian

    2013-06-01

    Medication errors account for 20% of medical errors in the United States with the largest risk at prescribing and administration. Analgesics or opioids are frequently used medications that can be associated with patient harm when prescribed or administered improperly. In an effort to decrease medication errors, Duke University Hospital implemented clinical decision support via computer provider order entry (CPOE) and "smart pump" technology, 2/2008, with the goal to decrease patient-controlled analgesia (PCA) adverse events. This project evaluated PCA safety events, reviewing voluntary report system and adverse drug events via surveillance (ADE-S), on intermediate and step-down units preimplementation and postimplementation of clinical decision support via CPOE and PCA smart pumps for the prescribing and administration of opioids therapy in the adult patient requiring analgesia for acute pain. Voluntary report system and ADE-S PCA events decreased based upon 1000 PCA days; ADE-S PCA events per 1000 PCA days decreased 22%, from 5.3 (pre) to 4.2 (post) (P = 0.09). Voluntary report system events decreased 72%, from 2.4/1000 PCA days (pre) to 0.66/1000 PCA days (post) and was statistically significant (P PCA events between time periods in both the ADE-S and voluntary report system data, thus supporting the recommendation of clinical decision support via CPOE and PCA smart pump technology.

  9. Social influence and perceptual decision making: a diffusion model analysis.

    Science.gov (United States)

    Germar, Markus; Schlemmer, Alexander; Krug, Kristine; Voss, Andreas; Mojzisch, Andreas

    2014-02-01

    Classic studies on social influence used simple perceptual decision-making tasks to examine how the opinions of others change individuals' judgments. Since then, one of the most fundamental questions in social psychology has been whether social influence can alter basic perceptual processes. To address this issue, we used a diffusion model analysis. Diffusion models provide a stochastic approach for separating the cognitive processes underlying speeded binary decisions. Following this approach, our study is the first to disentangle whether social influence on decision making is due to altering the uptake of available sensory information or due to shifting the decision criteria. In two experiments, we found consistent evidence for the idea that social influence alters the uptake of available sensory evidence. By contrast, participants did not adjust their decision criteria.

  10. Current status of safety analysis report for ANPP

    International Nuclear Information System (INIS)

    Amirjanyan, A.

    1999-01-01

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

  11. Hanford Site cleanup and transition: Risk data needs for decision making (Hanford risk data gap analysis decision guide)

    International Nuclear Information System (INIS)

    Gajewski, S.; Glantz, C.; Harper, B.; Bilyard, G.; Miller, P.

    1995-10-01

    Given the broad array of environmental problems, technical alternatives, and outcomes desired by different stakeholders at Hanford, DOE will have to make difficult resource allocations over the next few decades. Although some of these allocations will be driven purely by legal requirements, almost all of the major objectives of the cleanup and economic transition missions involve choices among alternative pathways. This study examined the following questions: what risk information is needed to make good decisions at Hanford; how do those data needs compare to the set(s) of risk data that will be generated by regulatory compliance activities and various non-compliance studies that are also concerned with risk? This analysis examined the Hanford Site missions, the Hanford Strategic Plan, known stakeholder values, and the most important decisions that have to be made at Hanford to determine a minimum domain of risk information required to make good decisions that will withstand legal, political, and technical scrutiny. The primary risk categories include (1) public health, (2) occupational health and safety, (3) ecological integrity, (4) cultural-religious welfare, and (5) socio-economic welfare

  12. Comparison of a Traditional Probabilistic Risk Assessment Approach with Advanced Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis L; Mandelli, Diego; Zhegang Ma

    2014-11-01

    As part of the Light Water Sustainability Program (LWRS) [1], the purpose of the Risk Informed Safety Margin Characterization (RISMC) [2] Pathway research and development (R&D) is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario. We also describe our approach we are using to represent advanced flooding analysis.

  13. A 'Toolbox' Equivalent Process for Safety Analysis Software

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Eng, Tony

    2004-01-01

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (Quality Assurance for Safety-Related Software) identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls that prevent or mitigate potential accidents. The development and maintenance of a collection, or 'toolbox', of multiple-site use, standard solution, Software Quality Assurance (SQA)-compliant safety software is one of the major improvements identified in the associated DOE Implementation Plan (IP). The DOE safety analysis toolbox will contain a set of appropriately quality-assured, configuration-controlled, safety analysis codes, recognized for DOE-broad, safety basis applications. Currently, six widely applied safety analysis computer codes have been designated for toolbox consideration. While the toolbox concept considerably reduces SQA burdens among DOE users of these codes, many users of unique, single-purpose, or single-site software may still have sufficient technical justification to continue use of their computer code of choice, but are thwarted by the multiple-site condition on toolbox candidate software. The process discussed here provides a roadmap for an equivalency argument, i.e., establishing satisfactory SQA credentials for single-site software that can be deemed ''toolbox-equivalent''. The process is based on the model established to meet IP Commitment 4.2.1.2: Establish SQA criteria for the safety analysis ''toolbox'' codes. Implementing criteria that establish the set of prescriptive SQA requirements are based on implementation plan/procedures from the Savannah River Site, also incorporating aspects of those from the Waste Isolation Pilot Plant (SNL component) and the Yucca Mountain Project. The major requirements are met with evidence of a software quality assurance plan, software requirements and design documentation, user's instructions, test report, a

  14. The adaptive safety analysis and monitoring system

    Science.gov (United States)

    Tu, Haiying; Allanach, Jeffrey; Singh, Satnam; Pattipati, Krishna R.; Willett, Peter

    2004-09-01

    The Adaptive Safety Analysis and Monitoring (ASAM) system is a hybrid model-based software tool for assisting intelligence analysts to identify terrorist threats, to predict possible evolution of the terrorist activities, and to suggest strategies for countering terrorism. The ASAM system provides a distributed processing structure for gathering, sharing, understanding, and using information to assess and predict terrorist network states. In combination with counter-terrorist network models, it can also suggest feasible actions to inhibit potential terrorist threats. In this paper, we will introduce the architecture of the ASAM system, and discuss the hybrid modeling approach embedded in it, viz., Hidden Markov Models (HMMs) to detect and provide soft evidence on the states of terrorist network nodes based on partial and imperfect observations, and Bayesian networks (BNs) to integrate soft evidence from multiple HMMs. The functionality of the ASAM system is illustrated by way of application to the Indian Airlines Hijacking, as modeled from open sources.

  15. Accident simulator development for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Amendola, A.; Mancini, G.

    1985-01-01

    This paper describes the basic features of a new concept of incident simulator, Response System Analyzed (RSA) which is being developed within the CEC JRC Research Program on Reactor Safety. Focusing on somewhat different aims than actual simulators, RSA development extends the field of application of simulators to the area of risk and reliability analysis and in particular to the identification of relevant sequences, to the modeling of human behavior and to the validation of operating procedures. The fundamental components of the project, i.e. the deterministic transient model of the plant, the automatic probabilistic driver and the human possible intervention modeling, are discussed in connection with the problem of their dynamic interaction. The analyses so far performed by separately testing RSA on significant study cases have shown encouraging results and have proven the feasibility of the overall program

  16. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  17. ARIES-RS safety design and analysis

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  18. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

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

    2005-03-01

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

  19. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  20. Influence of information about specific absorption rate (SAR) upon customers' purchase decisions and safety evaluation of mobile phones.

    Science.gov (United States)

    Wiedemann, Peter M; Schütz, Holger; Clauberg, Martin

    2008-02-01

    This study investigated whether the SAR value is a purchase-relevant characteristic of mobile phones for laypersons and what effect the disclosure of a precautionary SAR value has on laypersons' risk perception. The study consisted of two parts: Study part 1 used a conjoint analysis design to explore the relevance of the SAR value and other features of mobile phones for an intended buying decision. Study part 2 used an experimental, repeated measures design to examine the effect of the magnitude of SAR values and the disclosure of a precautionary SAR value on risk perception. In addition, the study included an analysis of prior concerns of the study participants with regard to mobile phone risks. Part 1 indicates that the SAR value has a high relevance for laypersons' purchase intentions. In the experimental purchase setting it ranks even before price and equipment features. The results of study part 2 show that providing information of a precautionary limit value does not influence risk perception. This result suggests that laypersons' underlying subjective "safety model" for mobile phones resembles more a "margin of safety" concept than a threshold concept. The latter observation holds true no matter how concerned the participants are. (c) 2007 Wiley-Liss, Inc.

  1. Fuzzy rationality and parameter elicitation in decision analysis

    Science.gov (United States)

    Nikolova, Natalia D.; Tenekedjiev, Kiril I.

    2010-07-01

    It is widely recognised by decision analysts that real decision-makers always make estimates in an interval form. An overview of techniques to find an optimal alternative among such with imprecise and interval probabilities is presented. Scalarisation methods are outlined as most appropriate. A proper continuation of such techniques is fuzzy rational (FR) decision analysis. A detailed representation of the elicitation process influenced by fuzzy rationality is given. The interval character of probabilities leads to the introduction of ribbon functions, whose general form and special cases are compared with the p-boxes. As demonstrated, approximation of utilities in FR decision analysis does not depend on the probabilities, but the approximation of probabilities is dependent on preferences.

  2. Safety analysis of a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

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

  3. Decision Analysis System for Selection of Appropriate Decontamination Technologies

    International Nuclear Information System (INIS)

    Ebadian, M.A.; Boudreaux, J.F.; Chinta, S.; Zanakis, S.H.

    1998-01-01

    The principal objective for designing Decision Analysis System for Decontamination (DASD) is to support DOE-EM's endeavor to employ the most efficient and effective technologies for treating radiologically contaminated surfaces while minimizing personnel and environmental risks. DASD will provide a tool for environmental decision makers to improve the quality, consistency, and efficacy of their technology selection decisions. The system will facilitate methodical comparisons between innovative and baseline decontamination technologies and aid in identifying the most suitable technologies for performing surface decontamination at DOE environmental restoration sites

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

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

  6. Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs

  7. Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs.

  8. Decision analysis for cleanup strategies in an urban environment

    International Nuclear Information System (INIS)

    Sinkko, K.; Ikaeheimonen, T.K.

    2000-01-01

    The values entering the decisions on protective actions, as concerning the society, are multidimensional. People have strong feelings and beliefs about these values, some of which are not numerically quantified and do not exist in monetary form. The decision analysis is applied in planning the recovery operations to clean up an urban environment in the event of a hypothetical nuclear power plant accident assisting in rendering explicit and apparent all factors involved and evaluating their relative importance. (author)

  9. Computerization of the safeguards analysis decision process

    International Nuclear Information System (INIS)

    Ehinger, M.H.

    1990-01-01

    This paper reports that safeguards regulations are evolving to meet new demands for timeliness and sensitivity in detecting the loss or unauthorized use of sensitive nuclear materials. The opportunities to meet new rules, particularly in bulk processing plants, involve developing techniques which use modern, computerized process control and information systems. Using these computerized systems in the safeguards analysis involves all the challenges of the man-machine interface experienced in the typical process control application and adds new dimensions to accuracy requirements, data analysis, and alarm resolution in the regulatory environment

  10. Preparing a Safety Analysis Report using the building block approach

    International Nuclear Information System (INIS)

    Herrington, C.C.

    1990-01-01

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

  11. The influence of sodium fires on LMFBRs safety analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-03-01

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

  12. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

    Justin, F.

    1979-01-01

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

  13. Employing Conjoint Analysis in Making Compensation Decisions.

    Science.gov (United States)

    Kienast, Philip; And Others

    1983-01-01

    Describes a method employing conjoint analysis that generates utility/cost ratios for various elements of the compensation package. Its superiority to simple preference surveys is examined. Results of a study of the use of this method in fringe benefit planning in a large financial institution are reported. (Author/JAC)

  14. Application of decision analysis in antibiotic formulary choices.

    Science.gov (United States)

    Szymusiak-Mutnick, B; Mutnick, A H

    1994-01-01

    To introduce the reader to the fundamentals involved in using decision analysis as a tool in evaluating the associated costs and effectiveness of comparable therapeutic agents. Currently available literature citations were used to provide the reader with basic references whose purpose is to provide a step-by-step approach for using Decision Analysis in conducting a cost-effective comparison of three commonly used antibiotics. Data were gathered from a previously conducted retrospective chart review where the three antibiotics were used for either prophylactic, empiric, or documented infections. Although this study was limited by its retrospective nature, the reader can use the data to appreciate the fundamentals of decision analysis. The continually changing climate in healthcare and the added visibility of pharmacologic agents in the treatment and prevention of disease has increased pressure on pharmacy departments to provide therapeutic agents that are cost-effective. Decision analysis can be used to compare therapeutic agents, in terms of financial as well as clinical outcomes, in a structured fashion that all members of the health care team can understand. The application of Decision analysis is appropriate for many therapeutic agents, not just antibiotics.

  15. Reliability Analysis for Safety Grade PLC(POSAFE-Q)

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. Criticality safety analysis for mockup facility

    International Nuclear Information System (INIS)

    Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

    2000-03-01

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

  17. Safety relief valve alternate analysis method

    International Nuclear Information System (INIS)

    Adams, R.H.; Javid, A.; Khatua, T.P.

    1981-01-01

    An experimental test program was started in the United States in 1976 to define and quantify Safety Relief Valve (SRV) phenomena in General Electric Mark I Suppression Chambers. The testing considered several discharged devices and was used to correlate SRV load prediction models. The program was funded by utilities with Mark I containments and has resulted in a detailed SRV load definition as a portion of the Mark I containment program Load Definition Report (LDR). The (USNRC) has reviewed and approved the LDR SRV load definition. In addition, the USNRC has permitted calibration of structural models used for predicting torus response to SRV loads. Model calibration is subject to confirmatory in-plant testing. The SRV methodology given in the LDR requires that transient dynamic pressures be applied to a torus structural model that includes a fluid added mass matrix. Preliminary evaluations of torus response have indicated order of magnitude conservatisms, with respect to test results, which could result in unrealistic containment modifications. In addition, structural response trends observed in full-scale tests between cold pipe, first valve actuation and hot pipe, subsequent valve actuation conditions have not been duplicated using current analysis methods. It was suggested by others that an energy approach using current fluid models be utilized to define loads. An alternate SRV analysis method is defined to correct suppression chamber structural response to a level that permits economical but conservative design. Simple analogs are developed for the purpose of correcting the analytical response obtained from LDR analysis methods. Analogs evaluated considered forced vibration and free vibration structural response. The corrected response correlated well with in-plant test response. The correlation of the analytical model at test conditions permits application of the alternate analysis method at design conditions. (orig./HP)

  18. Decision no. 2011-DC-0216 of the French nuclear safety authority from May 5, 2011, ordering the Laue Langevin Institute to proceed to a complementary safety evaluation of its basic nuclear facility (high flux reactor - INB no. 67) in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the Laue Langevin Institute, operator of the high flux research reactor (RHF) of Grenoble (France). (J.S.)

  19. Decision no. 2011-DC-0215 of the French nuclear safety authority from May 5, 2011, ordering ITER Organization to proceed to a complementary safety evaluation of its basic nuclear facility in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the ITER Organization, operator of the ITER tokamak facility of Cadarache (France). (J.S.)

  20. Decision no. 2011-DC-0222 of the French nuclear safety authority from May 5, 2011, ordering the Comurhex company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to Comurhex company, operator of the Tricastin uranium conversion plant (France). (J.S.)

  1. Decision no. 2011-DC-0223 of the French nuclear safety authority from May 5, 2011, ordering the MELOX SA company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to MELOX SA company, operator of the Melox MOX fuel fabrication plant of Marcoule (France). (J.S.)

  2. Decision no. 2011-DC-0218 of the French nuclear safety authority from May 5, 2011, ordering the EURODIF SA company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the EURODIF SA company, operator of the George Besse I uranium enrichment plant of the Tricastin site (France). (J.S.)

  3. Decision no. 2011-DC-0224 of the French nuclear safety authority from May 5, 2011, ordering the French atomic energy and alternative energies commission (CEA) to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the French atomic energy commission (CEA). (J.S.)

  4. Decision no. 2011-DC-0214 of the French nuclear safety authority from May 5, 2011, ordering CIS bio international company to proceed to a complementary safety evaluation of its basic nuclear facility in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to CIS bio international company, operator of the radiopharmaceuticals fabrication facility (INB 29) of Saclay (France). (J.S.)

  5. Decision no. 2011-DC-0219 of the French nuclear safety authority from May 5, 2011, ordering the SOCATRI company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2011-01-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the SOCATRI company, operator of the nuclear dismantling and waste processing plants of the Tricastin site (France). (J.S.)

  6. Safety evaluation by living probabilistic safety assessment. Procedures and applications for planning of operational activities and analysis of operating experience

    International Nuclear Information System (INIS)

    Johanson, Gunnar; Holmberg, J.

    1994-01-01

    Living Probabilistic Safety Assessment (PSA) is a daily safety management system and it is based on a plant-specific PSA and supporting information systems. In the living use of PSA, plant status knowledge is used to represent actual plant safety status in monitoring or follow-up perspective. The PSA model must be able to express the risk at a given time and plant configuration. The process, to update the PSA model to represent the current or planned configuration and to use the model to evaluate and direct the changes in the configuration, is called living PSA programme. The main purposes to develop and increase the usefulness of living PSA are: Long term safety planning: To continue the risk assessment process started with the basic PSA by extending and improving the basic models and data to provide a general risk evaluation tool for analyzing the safety effects of changes in plant design and procedures. Risk planning of operational activities: To support the operational management by providing means for searching optimal operational maintenance and testing strategies from the safety point of view. The results provide support for risk decision making in the short term or in a planning mode. The operational limits and conditions given by technical specifications can be analyzed by evaluating the risk effects of alternative requirements in order to balance the requirements with respect to operational flexibility and plant economy. Risk analysis of operating experience: To provide a general risk evaluation tool for analyzing the safety effects of incidents and plant status changes. The analyses are used to: identify possible high risk situations, rank the occurred events from safety point of view, and get feedback from operational events for the identification of risk contributors. This report describes the methods, models and applications required to continue the process towards a living use of PSA. 19 tabs, 20 figs

  7. Making nuclear power plant operational decisions using probabilistic safety assessment information and personal computers. Working material

    International Nuclear Information System (INIS)

    1991-01-01

    PRISIM described in this case study makes a PSA useful to decision makers like plant managers, operational personnel or safety assessors because it provides a rapid access to specific information and the ability to generate updated PSA results that reflect the plant status at a particular time. From the capabilities of PRISIM one can conclude that the ability of a user friendly update of the system model in the PC or changes in the data files at the computer is not realized to data. Also the calculation of averaged probabilities instead of time dependent instantaneous probabilities is a sort of a restriction and will be changed in the future. 5 refs, 34 figs, 3 tabs

  8. Safety analysis and synthesis using fuzzy sets and evidential reasoning

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  9. TESTING MULTI-CRITERIA DECISION ANALYSIS FOR MORE TRANSPARENT RESOURCE-ALLOCATION DECISION MAKING IN COLOMBIA.

    Science.gov (United States)

    Castro Jaramillo, Hector Eduardo; Goetghebeur, Mireille; Moreno-Mattar, Ornella

    2016-01-01

    In 2012, Colombia experienced an important institutional transformation after the establishment of the Health Technology Assessment Institute (IETS), the disbandment of the Regulatory Commission for Health and the reassignment of reimbursement decision-making powers to the Ministry of Health and Social Protection (MoHSP). These dynamic changes provided the opportunity to test Multi-Criteria Decision Analysis (MCDA) for systematic and more transparent resource-allocation decision-making. During 2012 and 2013, the MCDA framework Evidence and Value: Impact on Decision Making (EVIDEM) was tested in Colombia. This consisted of a preparatory stage in which the investigators conducted literature searches and produced HTA reports for four interventions of interest, followed by a panel session with decision makers. This method was contrasted with a current approach used in Colombia for updating the publicly financed benefits package (POS), where narrative health technology assessment (HTA) reports are presented alongside comprehensive budget impact analyses (BIAs). Disease severity, size of population, and efficacy ranked at the top among fifteen preselected relevant criteria. MCDA estimates of technologies of interest ranged between 71 to 90 percent of maximum value. The ranking of technologies was sensitive to the methods used. Participants considered that a two-step approach including an MCDA template, complemented by a detailed BIA would be the best approach to assist decision-making in this context. Participants agreed that systematic priority setting should take place in Colombia. This work may serve as the basis to the MoHSP on its interest of setting up a systematic and more transparent process for resource-allocation decision-making.

  10. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. Multi-criteria decision analysis for use in transport decision making

    DEFF Research Database (Denmark)

    the recent years that besides the social costs and benefits associated with transport other impacts that are more difficult to monetise should also have influence on the decision making process. This is in many developed countries realised in the transport planning, which takes into account a wide range......, however, commonly agreed that the final decision making concerning transport infrastructure projects in many cases will depend on other aspects besides the monetary ones assessed in a socio-economic analysis. Nevertheless, an assessment framework such as the Danish one (DMT, 2003) does not provide any...... specific guidelines on how to include the strategic impacts; it merely suggests describing the impacts verbally and keeping them in mind during the decision process. A coherent, well-structured, flexible, straight forward evaluation method, taking into account all the requirements of a transport...

  12. Selection of safety officers in an indian construction organization by using grey relational analysis

    Directory of Open Access Journals (Sweden)

    Sunku Venkata Siva Rajaprasad

    2018-03-01

    Full Text Available Stakeholders are responsible for implementing the occupational health and safety provisions in an organization. Irrespective of organization, the role of safety department is purely advisory as it coordinates with all the departments, and this is crucial to improve the performance. Selection of safety officer is vital job for any organization; it should not only be based on qualifications of the applicant, the incumbent should also have sufficient exposure in implementing proactive measures. The process of selection is complex and choosing the right safety professional is a vital decision. The safety performance of an organization relies on the systems being implemented by the safety officer. Application of multi criteria decision-making tools is helpful as a selection process. The present study proposes the grey relational analysis(GRA for selection of the safety officers in an Indian construction organization. This selection method considers fourteen criteria appropriate to the organization and has ranked the results. The data was also analyzed by using technique for order Preference by Similarity to an Ideal solution (TOPSIS and results of both the methods are strongly correlated

  13. Decision-making in irrigation networks: Selecting appropriate canal structures using multi-attribute decision analysis.

    Science.gov (United States)

    Hosseinzade, Zeinab; Pagsuyoin, Sheree A; Ponnambalam, Kumaraswamy; Monem, Mohammad J

    2017-12-01

    The stiff competition for water between agriculture and non-agricultural production sectors makes it necessary to have effective management of irrigation networks in farms. However, the process of selecting flow control structures in irrigation networks is highly complex and involves different levels of decision makers. In this paper, we apply multi-attribute decision making (MADM) methodology to develop a decision analysis (DA) framework for evaluating, ranking and selecting check and intake structures for irrigation canals. The DA framework consists of identifying relevant attributes for canal structures, developing a robust scoring system for alternatives, identifying a procedure for data quality control, and identifying a MADM model for the decision analysis. An application is illustrated through an analysis for automation purposes of the Qazvin irrigation network, one of the oldest and most complex irrigation networks in Iran. A survey questionnaire designed based on the decision framework was distributed to experts, managers, and operators of the Qazvin network and to experts from the Ministry of Power in Iran. Five check structures and four intake structures were evaluated. A decision matrix was generated from the average scores collected from the survey, and was subsequently solved using TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution) method. To identify the most critical structure attributes for the selection process, optimal attribute weights were calculated using Entropy method. For check structures, results show that the duckbill weir is the preferred structure while the pivot weir is the least preferred. Use of the duckbill weir can potentially address the problem with existing Amil gates where manual intervention is required to regulate water levels during periods of flow extremes. For intake structures, the Neyrpic® gate and constant head orifice are the most and least preferred alternatives, respectively. Some advantages

  14. Benefit-Risk Analysis for Decision-Making: An Approach.

    Science.gov (United States)

    Raju, G K; Gurumurthi, K; Domike, R

    2016-12-01

    The analysis of benefit and risk is an important aspect of decision-making throughout the drug lifecycle. In this work, the use of a benefit-risk analysis approach to support decision-making was explored. The proposed approach builds on the qualitative US Food and Drug Administration (FDA) approach to include a more explicit analysis based on international standards and guidance that enables aggregation and comparison of benefit and risk on a common basis and a lifecycle focus. The approach is demonstrated on six decisions over the lifecycle (e.g., accelerated approval, withdrawal, and traditional approval) using two case studies: natalizumab for multiple sclerosis (MS) and bedaquiline for multidrug-resistant tuberculosis (MDR-TB). © 2016 American Society for Clinical Pharmacology and Therapeutics.

  15. Big Data Risk Analysis for Rail Safety?

    OpenAIRE

    Van Gulijk, Coen; Hughes, Peter; Figueres-Esteban, Miguel; Dacre, Marcus; Harrison, Chris; HUD; RSSB

    2015-01-01

    Computer scientists believe that the enormous amounts of data in the internet will unchain a management revolution of uncanny proportions. Yet, to date, the potential benefit of this revolution is scantily investigated for safety and risk management. This paper gives a brief overview of a research programme that investigates how the new internet-driven data-revolution could benefit safety and risk management for railway safety in the UK. The paper gives a brief overview the current activities...

  16. Safety management - policy, analysis and implementation

    International Nuclear Information System (INIS)

    Allen, F.R.

    1993-01-01

    The nuclear industry is moving towards a period of ever increasing emphasis on business performance and profitability. Safety has, of course, always been a major concern of management in the nuclear industry and elsewhere. The civil aviation industry , for example, has had a similar concern for safety. Other industry sectors are also developing safety management as a response to events within and outside their sectors. In this paper the way that the risk management process as a whole is being addressed is looked at. Can we use risk management, initially a safety-orientated tool, to improve business performance? (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  18. Closed-Loop Analysis of Soft Decisions for Serial Links

    Science.gov (United States)

    Lansdowne, Chatwin A.; Steele, Glen F.; Zucha, Joan P.; Schlesinger, Adam M.

    2013-01-01

    We describe the benefit of using closed-loop measurements for a radio receiver paired with a counterpart transmitter. We show that real-time analysis of the soft decision output of a receiver can provide rich and relevant insight far beyond the traditional hard-decision bit error rate (BER) test statistic. We describe a Soft Decision Analyzer (SDA) implementation for closed-loop measurements on single- or dual- (orthogonal) channel serial data communication links. The analyzer has been used to identify, quantify, and prioritize contributors to implementation loss in live-time during the development of software defined radios. This test technique gains importance as modern receivers are providing soft decision symbol synchronization as radio links are challenged to push more data and more protocol overhead through noisier channels, and software-defined radios (SDRs) use error-correction codes that approach Shannon's theoretical limit of performance.

  19. Multi-criteria decision analysis and environmental risk assessment for nanomaterials

    International Nuclear Information System (INIS)

    Linkov, Igor; Satterstrom, F. Kyle; Steevens, Jeffery; Ferguson, Elizabeth; Pleus, Richard C.

    2007-01-01

    Nanotechnology is a broad and complex discipline that holds great promise for innovations that can benefit mankind. Yet, one must not overlook the wide array of factors involved in managing nanomaterial development, ranging from the technical specifications of the material to possible adverse effects in humans. Other opportunities to evaluate benefits and risks are inherent in environmental health and safety (EHS) issues related to nanotechnology. However, there is currently no structured approach for making justifiable and transparent decisions with explicit trade-offs between the many factors that need to be taken into account. While many possible decision-making approaches exist, we believe that multi-criteria decision analysis (MCDA) is a powerful and scientifically sound decision analytical framework for nanomaterial risk assessment and management. This paper combines state-of-the-art research in MCDA methods applicable to nanotechnology with a hypothetical case study for nanomaterial management. The example shows how MCDA application can balance societal benefits against unintended side effects and risks, and how it can also bring together multiple lines of evidence to estimate the likely toxicity and risk of nanomaterials given limited information on physical and chemical properties. The essential contribution of MCDA is to link this performance information with decision criteria and weightings elicited from scientists and managers, allowing visualization and quantification of the trade-offs involved in the decision-making process

  20. Multi-criteria decision analysis and environmental risk assessment for nanomaterials

    Science.gov (United States)

    Linkov, Igor; Satterstrom, F. Kyle; Steevens, Jeffery; Ferguson, Elizabeth; Pleus, Richard C.

    2007-08-01

    Nanotechnology is a broad and complex discipline that holds great promise for innovations that can benefit mankind. Yet, one must not overlook the wide array of factors involved in managing nanomaterial development, ranging from the technical specifications of the material to possible adverse effects in humans. Other opportunities to evaluate benefits and risks are inherent in environmental health and safety (EHS) issues related to nanotechnology. However, there is currently no structured approach for making justifiable and transparent decisions with explicit trade-offs between the many factors that need to be taken into account. While many possible decision-making approaches exist, we believe that multi-criteria decision analysis (MCDA) is a powerful and scientifically sound decision analytical framework for nanomaterial risk assessment and management. This paper combines state-of-the-art research in MCDA methods applicable to nanotechnology with a hypothetical case study for nanomaterial management. The example shows how MCDA application can balance societal benefits against unintended side effects and risks, and how it can also bring together multiple lines of evidence to estimate the likely toxicity and risk of nanomaterials given limited information on physical and chemical properties. The essential contribution of MCDA is to link this performance information with decision criteria and weightings elicited from scientists and managers, allowing visualization and quantification of the trade-offs involved in the decision-making process.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

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

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

    International Nuclear Information System (INIS)

    2011-01-01

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

  3. Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

    Directory of Open Access Journals (Sweden)

    Zongzhi Li

    2017-04-01

    Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

  4. The cost of harm and savings through safety: using simulated patients for leadership decision support.

    Science.gov (United States)

    Denham, Charles R; Guilloteau, Franck R

    2012-09-01

    The ultimate objective of this program is to provide an approach to understanding and communicating health-care harm and cost to compel health-care provider leadership teams to vote "yes" to investments in patient safety initiatives, with the confidence that clinical, financial, and operational performance will be improved by such programs. Through a coordinated combination of literature evaluations, careful mapping of high impact scenarios using simulated patients and consensus review of clinical, operational, and financial factors, we confirmed value in such approaches to decision support information for hospital leadership teams to invest in patient safety projects. The study resulted in the following preliminary findings: ·Communication between hospital quality and finance departments can be much improved by direct collaborative relationships through regular meetings to help both clarify direct costs, indirect costs, and the savings of waste and harm to patients by avoidance of infections. ·Governance leaders and the professional administrative leaders should consider establishing the structures and systems necessary to act on risks and hazards as they evolve to deploy resources to areas of harm and risk. ·Quality and Infection Control Professionals can best wage their war on healthcare waste and harm by keeping abreast of the latest literature regarding the latest measures, standards, and safe practices for healthcare-acquired infections and hospital-acquired conditions. ·Regular reviews of patients with health-careYassociated infections, with direct attention to the attributable cost of treatment and how financial waste and harm to patients may be avoided, may provide hospital leaders with new insights for improvement. ·If hospitals developed their own risk scenarios to determine impact of harm and waste from hospital-acquired conditions in addition to impact scenarios for specific processes through technology and process innovations, they would have

  5. Use of decision analysis techniques to determine Hanford cleanup priorities

    International Nuclear Information System (INIS)

    Fassbender, L.; Gregory, R.; Winterfeldt, D. von; John, R.

    1992-01-01

    In January 1991, the U.S. Department of Energy (DOE) Richland Field Office, Westinghouse Hanford Company, and the Pacific Northwest Laboratory initiated the Hanford Integrated Planning Process (HIPP) to ensure that technically sound and publicly acceptable decisions are made that support the environmental cleanup mission at Hanford. One of the HIPP's key roles is to develop an understanding of the science and technology (S and T) requirements to support the cleanup mission. This includes conducting an annual systematic assessment of the S and T needs at Hanford to support a comprehensive technology development program and a complementary scientific research program. Basic to success is a planning and assessment methodology that is defensible from a technical perspective and acceptable to the various Hanford stakeholders. Decision analysis techniques were used to help identify and prioritize problems and S and T needs at Hanford. The approach used structured elicitations to bring many Hanford stakeholders into the process. Decision analysis, which is based on the axioms and methods of utility and probability theory, is especially useful in problems characterized by uncertainties and multiple objectives. Decision analysis addresses uncertainties by laying out a logical sequence of decisions, events, and consequences and by quantifying event and consequence probabilities on the basis of expert judgments

  6. An Analysis of Laboratory Safety in Texas.

    Science.gov (United States)

    Fuller, Edward J.; Picucci, Ali Callicoatte; Collins, James W.; Swann, Philip

    This paper reports on a survey to discover the types of laboratory accidents that occur in Texas public schools, the factors associated with such accidents, and the practices of schools with regard to current laboratory safety requirements. The purpose of the survey is to better understand safety conditions in Texas public schools and to help…

  7. Quantification of a decision-making failure probability of the accident management using cognitive analysis model

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Ohtani, Masanori; Fujita, Yushi

    2002-01-01

    In the nuclear power plant, much knowledge is acquired through probabilistic safety assessment (PSA) of a severe accident, and accident management (AM) is prepared. It is necessary to evaluate the effectiveness of AM using the decision-making failure probability of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipments in PSA. However, there has been no suitable qualification method for PSA so far to obtain the decision-making failure probability, because the decision-making failure of an emergency organization treats the knowledge based error. In this work, we developed a new method for quantification of the decision-making failure probability of an emergency organization using cognitive analysis model, which decided an AM strategy, in a nuclear power plant at the severe accident, and tried to apply it to a typical pressurized water reactor (PWR) plant. As a result: (1) It could quantify the decision-making failure probability adjusted to PSA for general analysts, who do not necessarily possess professional human factors knowledge, by choosing the suitable value of a basic failure probability and an error-factor. (2) The decision-making failure probabilities of six AMs were in the range of 0.23 to 0.41 using the screening evaluation method and in the range of 0.10 to 0.19 using the detailed evaluation method as the result of trial evaluation based on severe accident analysis of a typical PWR plant, and a result of sensitivity analysis of the conservative assumption, failure probability decreased about 50%. (3) The failure probability using the screening evaluation method exceeded that using detailed evaluation method by 99% of probability theoretically, and the failure probability of AM in this study exceeded 100%. From this result, it was shown that the decision-making failure probability was more conservative than the detailed evaluation method, and the screening evaluation method satisfied

  8. Quantification of a decision-making failure probability of the accident management using cognitive analysis model

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Yoshitaka; Ohtani, Masanori [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Fujita, Yushi [TECNOVA Corp., Tokyo (Japan)

    2002-09-01

    In the nuclear power plant, much knowledge is acquired through probabilistic safety assessment (PSA) of a severe accident, and accident management (AM) is prepared. It is necessary to evaluate the effectiveness of AM using the decision-making failure probability of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipments in PSA. However, there has been no suitable qualification method for PSA so far to obtain the decision-making failure probability, because the decision-making failure of an emergency organization treats the knowledge based error. In this work, we developed a new method for quantification of the decision-making failure probability of an emergency organization using cognitive analysis model, which decided an AM strategy, in a nuclear power plant at the severe accident, and tried to apply it to a typical pressurized water reactor (PWR) plant. As a result: (1) It could quantify the decision-making failure probability adjusted to PSA for general analysts, who do not necessarily possess professional human factors knowledge, by choosing the suitable value of a basic failure probability and an error-factor. (2) The decision-making failure probabilities of six AMs were in the range of 0.23 to 0.41 using the screening evaluation method and in the range of 0.10 to 0.19 using the detailed evaluation method as the result of trial evaluation based on severe accident analysis of a typical PWR plant, and a result of sensitivity analysis of the conservative assumption, failure probability decreased about 50%. (3) The failure probability using the screening evaluation method exceeded that using detailed evaluation method by 99% of probability theoretically, and the failure probability of AM in this study exceeded 100%. From this result, it was shown that the decision-making failure probability was more conservative than the detailed evaluation method, and the screening evaluation method satisfied

  9. Analysis on safety production in coal mines Henan Province

    Institute of Scientific and Technical Information of China (English)

    KONG Liu-an; ZHANG Wen-yong

    2006-01-01

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

  10. Safety Injection Tank Performance Analysis Using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Oan; Lee, Jeong Ik; Nietiadi Yohanes Setiawan [KAIST, Daejeon (Korea, Republic of); Addad Yacine [KUSTAR, Abu Dhabi (United Arab Emirates); Bang, Young Seok; Yoo, Seung Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    This may affect the core cooling capability and threaten the fuel integrity during LOCA situations. However, information on the nitrogen flow rate during discharge is very limited due to the associated experimental measurement difficulties, and these phenomena are hardly reflected in current 1D system codes. In the current study, a CFD analysis is presented which hopefully should allow obtaining a more realistic prediction of the SIT performance which can then be reflected on 1D system codes to simulate various accident scenarios. Current Computational Fluid Dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study aims to find a better CFD prediction and more accurate modeling to predict the system performance during accident scenarios. The safety injection tank with fluidic device was analyzed using commercial CFD. A fine resolution grid was used to capture the vortex of the fluidic device. The calculation so far has shown good consistency with the experiment. Calculation should complete by the conference date and will be thoroughly analyzed to be discussed. Once a detailed CFD computation is finished, a small-scale experiment will be conducted for the given conditions. Using the experimental results and the CFD model, physical models can be validated to give more reliable results. The data from CFD and experiments will provide a more accurate K-factor of the fluidic device which can later be applied in system code inputs.

  11. Making Good Decisions in Healthcare with Multi-Criteria Decision Analysis: The Use, Current Research and Future Development of MCDA.

    Science.gov (United States)

    Mühlbacher, Axel C; Kaczynski, Anika

    2016-02-01

    Healthcare decision making is usually characterized by a low degree of transparency. The demand for transparent decision processes can be fulfilled only when assessment, appraisal and decisions about health technologies are performed under a systematic construct of benefit assessment. The benefit of an intervention is often multidimensional and, thus, must be represented by several decision criteria. Complex decision problems require an assessment and appraisal of various criteria; therefore, a decision process that systematically identifies the best available alternative and enables an optimal and transparent decision is needed. For that reason, decision criteria must be weighted and goal achievement must be scored for all alternatives. Methods of multi-criteria decision analysis (MCDA) are available to analyse and appraise multiple clinical endpoints and structure complex decision problems in healthcare decision making. By means of MCDA, value judgments, priorities and preferences of patients, insurees and experts can be integrated systematically and transparently into the decision-making process. This article describes the MCDA framework and identifies potential areas where MCDA can be of use (e.g. approval, guidelines and reimbursement/pricing of health technologies). A literature search was performed to identify current research in healthcare. The results showed that healthcare decision making is addressing the problem of multiple decision criteria and is focusing on the future development and use of techniques to weight and score different decision criteria. This article emphasizes the use and future benefit of MCDA.

  12. Compositional Safety Analysis using Barrier Certificates

    DEFF Research Database (Denmark)

    Sloth, Christoffer; Pappas, George J.; Wisniewski, Rafael

    2012-01-01

    This paper proposes a compositional method for verifying the safety of a dynamical system, given as an interconnection of subsystems. The safety verification is conducted by the use of the barrier certificate method; hence, the contribution of this paper is to show how to obtain compositional...... conditions for safety verification. We show how to formulate the verification problem, as a composition of coupled subproblems, each given for one subsystem. Furthermore, we show how to find the compositional barrier certificates via linear and sum of squares programming problems. The proposed method makes...... it possible to verify the safety of higher dimensional systems, than the method for centrally computed barrier certificates. This is demonstrated by verifying the safety of an emergency shutdown of a wind turbine....

  13. Using Cognitive Work Analysis to fit decision support tools to nurse managers' work flow.

    Science.gov (United States)

    Effken, Judith A; Brewer, Barbara B; Logue, Melanie D; Gephart, Sheila M; Verran, Joyce A

    2011-10-01

    To better understand the environmental constraints on nurse managers that impact their need for and use of decision support tools, we conducted a Cognitive Work Analysis (CWA). A complete CWA includes system analyses at five levels: work domain, decision-making procedures, decision-making strategies, social organization/collaboration, and worker skill level. Here we describe the results of the Work Domain Analysis (WDA) portion in detail then integrate the WDA with other portions of the CWA, reported previously, to generate a more complete picture of the nurse manager's work domain. Data for the WDA were obtained from semi-structured interviews with nurse managers, division directors, CNOs, and other managers (n = 20) on 10 patient care units in three Arizona hospitals. The WDA described the nurse manager's environment in terms of the constraints it imposes on the nurse manager's ability to achieve targeted outcomes through organizational goals and priorities, functions, processes, as well as work objects and resources (e.g., people, equipment, technology, and data). Constraints were identified and summarized through qualitative thematic analysis. The results highlight the competing priorities, and external and internal constraints that today's nurse managers must satisfy as they try to improve quality and safety outcomes on their units. Nurse managers receive a great deal of data, much in electronic format. Although dashboards were perceived as helpful because they integrated some data elements, no decision support tools were available to help nurse managers with planning or answering "what if" questions. The results suggest both the need for additional decision support to manage the growing complexity of the environment, and the constraints the environment places on the design of that technology if it is to be effective. Limitations of the study include the small homogeneous sample and the reliance on interview data targeting safety and quality. Copyright © 2011

  14. Analysis for making a regulatory decision to equipment of industrial gammagraphy in Argentin

    International Nuclear Information System (INIS)

    Ermacora, Marcela G.; Vidal, Dora N.; Alonso, Maria T.

    2013-01-01

    Industrial gammagraphy is a practice widely used as a nondestructive testing technique in Argentina. Experience worldwide has shown the need for an improvement in the intrinsic safety of the equipment used in this lab. In response to this reason, the board of the Nuclear Regulatory Authority (ARN) has considered a proposal to withdraw service movement and much of the equipment inventory scan belonging to industrial facilities nationwide. The main objective of this paper is to present the results of the analysis performed to support the above proposal. The main elements of evaluation can be summarized as follows: I) the teams that do not conform to international recommendations regarding compliance with key safety requirements of international standards such as ISO 3999:2004 (E) R adiation protection - Industrial Apparatus for gamma radiography - Specifications for performance, design and tests ; II) the decision by some manufacturers to discontinue production of certain models of equipment and the provision of spare parts, and III) the validity of certificates bulk type B (U) for transport. In conclusion, it highlights the importance of a regulatory decision supplementary to the Standard AR 7.9.1 concerning the operation of scan equipment industry, based on current international recommendations and Argentina's commitment to good practice and safety culture which can lead to a positive impact on radiation safety in this art

  15. Safety analysis of Oi nuclear power plant

    International Nuclear Information System (INIS)

    1979-01-01

    The transient phenomena in Oi nuclear power plant were analyzed, especially on the water level fluctuation and the capability of natural circulation in the primary loop, under the assumptions that the feed water for steam generators is totally lost, and the relief valve on the pressurizer, which is actuated due to the pressure rise in the primary system, is stuck and kept open. These assumptions are related to the TMI accident. The analysing conditions are 1) the main feed water flow is totally lost suddenly during the rated power operation of the reactor, 2) two motor-driven auxiliary feed water pumps are started manually fifteen minutes after the accident initiation, 3) one relief valve on the pressurizer is opened fifteen seconds after the accident initiation and kept open, 4) the reactor is scrammed thirty three seconds after the accident initiation, 5) the turbine is tripped 33.5 seconds after the accident initiation, etc. Two cases were analysed, namely 3,800 seconds and 1,200 seconds after the accident initiation. The analytical code RELEP4/Mod5/U2/J1 was utilized for this analysis. The level fluctuation in the pressurizer after the accident initiation, the flow rate fluctuation through the pressurizer relief valve, especially that of steam, liquid single phase and two phase flows, the water level in the upper plenum in the pressure vessel, the change of flow rate at core inlet, the average pressure in the core, and the temperature fluctuation of coolant in the core, the variation of void fraction in the core, and the change of surface temperature of fuel rods are presented as the analysis results, and they are evaluated. It is recognized that the plant safety is kept under the assumed accident conditions in the Oi nuclear power plant. (Nakai, Y.)

  16. Software FMEA analysis for safety-related application software

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

    2012-04-01

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

  18. A decision analysis of an exploratory studies facility

    International Nuclear Information System (INIS)

    Merkhofer, M.W.; Gnirk, P.

    1992-01-01

    This paper reports that an Exploratory Studied Facility (ESF) is planned to support the characterization of a potential site for a high-level nuclear waste repository at Yucca Mountain, NV. The selection of a design for the ESF is a critical characterization decision because the ESF design may affect the accuracy of characterization testing an constrains subsequent repository design. To assist the design process, a comparative evaluation was conducted to rank 34 alternative ESF-repository designs. The evaluation relied on techniques from formal decision analysis, including decision trees and multiattribute utility analysis (MUA). The results helped to identify favorable design features and enabled the Department of Energy to adopt an improved ESF design

  19. Using real options analysis to support strategic management decisions

    Science.gov (United States)

    Kabaivanov, Stanimir; Markovska, Veneta; Milev, Mariyan

    2013-12-01

    Decision making is a complex process that requires taking into consideration multiple heterogeneous sources of uncertainty. Standard valuation and financial analysis techniques often fail to properly account for all these sources of risk as well as for all sources of additional flexibility. In this paper we explore applications of a modified binomial tree method for real options analysis (ROA) in an effort to improve decision making process. Usual cases of use of real options are analyzed with elaborate study on the applications and advantages that company management can derive from their application. A numeric results based on extending simple binomial tree approach for multiple sources of uncertainty are provided to demonstrate the improvement effects on management decisions.

  20. Moon manned missions radiation safety analysis

    Science.gov (United States)

    Tripathi, R. K.; Wilson, J. W.; de Anlelis, G.; Badavi, F. F.

    , from very simple shelters to more complex bases, are considered in full detail (e.g., shape, thickness, materials, etc) with considerations of various shielding strategies. In this first analysis all the shape considered are cylindrical or composed of combination of cylinders. Moreover, a radiation safety analysis of more future possible habitats like lava tubes has been also performed.

  1. Holistic safety analysis for advanced nuclear power plants

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. Safety analysis of the nuclear chemistry Building 151

    International Nuclear Information System (INIS)

    Kvam, D.

    1984-01-01

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

  3. Special characteristics of the safety analysis of HWRs

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

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

  4. An extended data envelopment analysis for the decision-making

    Directory of Open Access Journals (Sweden)

    Xiao-Li Meng

    2017-10-01

    Full Text Available Abstract Based on the CCR model, we propose an extended data envelopment analysis to evaluate the efficiency of decision making units with historical input and output data. The contributions of the work are threefold. First, the input and output data of the evaluated decision making unit are variable over time, and time series method is used to analyze and predict the data. Second, there are many sample decision making units, which are divided into several ordered sample standards in terms of production strategy, and the constraint condition consists of one of the sample standards. Furthermore, the efficiency is illustrated by considering the efficiency relationship between the evaluated decision making unit and sample decision making units from constraint condition. Third, to reduce the computation complexity, we introduce an algorithm based on the binary search tree in the model to choose the sample standard that has similar behavior with the evaluated decision making unit. Finally, we provide two numerical examples to illustrate the proposed model.

  5. STARS software tool for analysis of reliability and safety

    International Nuclear Information System (INIS)

    Poucet, A.; Guagnini, E.

    1989-01-01

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

  6. Transit safety & security statistics & analysis 2002 annual report (formerly SAMIS)

    Science.gov (United States)

    2004-12-01

    The Transit Safety & Security Statistics & Analysis 2002 Annual Report (formerly SAMIS) is a compilation and analysis of mass transit accident, casualty, and crime statistics reported under the Federal Transit Administrations (FTAs) National Tr...

  7. Transit safety & security statistics & analysis 2003 annual report (formerly SAMIS)

    Science.gov (United States)

    2005-12-01

    The Transit Safety & Security Statistics & Analysis 2003 Annual Report (formerly SAMIS) is a compilation and analysis of mass transit accident, casualty, and crime statistics reported under the Federal Transit Administrations (FTAs) National Tr...

  8. A Framework for an Integrated Risk Informed Decision Making Process. INSAG-25. A Report by the International Nuclear Safety Group

    International Nuclear Information System (INIS)

    2011-01-01

    There is general international agreement, as reflected in various IAEA Safety Standards on nuclear reactor design and operation, that both deterministic and probabilistic analyses contribute to reactor safety by providing insights, perspective, comprehension and balance. Accordingly, the integration of deterministic and probabilistic analyses is increasing to support design, safety evaluation and operations. Additionally, application of these approaches to physical security is now being considered by several Member States. Deterministic and probabilistic analyses yield outputs that are complementary to each other. There is thus a need to use a structured framework for consideration of deterministic and probabilistic techniques and findings. In this process, it is appropriate to encourage a balance between deterministic approaches, probabilistic analyses and other factors (see Section 3) in order to achieve an integrated decision making process that serves in an optimal fashion to ensure nuclear reactor safety. This report presents such a framework - a framework that is termed 'integrated risk informed decision making' (IRIDM). While the details of IRIDM methods may change with better understanding of the subject, the framework presented in this report is expected to apply for the foreseeable future. IRIDM depends on the integration of a wide variety of information, insights and perspectives, as well as the commitment of designers, operators and regulatory authorities to use risk information in their decisions. This report thus focuses on key IRIDM aspects, as well considerations that bear on their application which should be taken into account in order to arrive at sound risk informed decisions. This report is intended to be in harmony with the IAEA Safety Standards and various INSAG reports relating to safety assessment and verification, and seeks to convey an appropriate approach to enhance nuclear reactor safety

  9. A Framework for an Integrated Risk Informed Decision Making Process. INSAG-25. A Report by the International Nuclear Safety Group

    International Nuclear Information System (INIS)

    2014-01-01

    There is general international agreement, as reflected in various IAEA Safety Standards on nuclear reactor design and operation, that both deterministic and probabilistic analyses contribute to reactor safety by providing insights, perspective, comprehension and balance. Accordingly, the integration of deterministic and probabilistic analyses is increasing to support design, safety evaluation and operations. Additionally, application of these approaches to physical security is now being considered by several Member States. Deterministic and probabilistic analyses yield outputs that are complementary to each other. There is thus a need to use a structured framework for consideration of deterministic and probabilistic techniques and findings. In this process, it is appropriate to encourage a balance between deterministic approaches, probabilistic analyses and other factors (see Section 3) in order to achieve an integrated decision making process that serves in an optimal fashion to ensure nuclear reactor safety. This report presents such a framework - a framework that is termed 'integrated risk informed decision making' (IRIDM). While the details of IRIDM methods may change with better understanding of the subject, the framework presented in this report is expected to apply for the foreseeable future. IRIDM depends on the integration of a wide variety of information, insights and perspectives, as well as the commitment of designers, operators and regulatory authorities ers, operators and regulatory authorities to use risk information in their decisions. This report thus focuses on key IRIDM aspects, as well considerations that bear on their application which should be taken into account in order to arrive at sound risk informed decisions. This report is intended to be in harmony with the IAEA Safety Standards and various INSAG reports relating to safety assessment and verification, and seeks to convey an appropriate approach to enhance nuclear reactor safety

  10. Safety analysis report 231-Z Building

    Energy Technology Data Exchange (ETDEWEB)

    Powers, C.S.

    1989-03-01

    This report provides an intensive review of the nuclear safety of the operation of the 231-Z Building. For background information complete descriptions of the floor plan, building services, alarm systems, and glove box systems are included in this report. In addition, references are included to The Plutonium Laboratory Radiation Work Procedures, Safety Guides, 231-Z Operating Procedures Manual and Nuclear Materials accountability Procedures. Engineered and administrative features contribute to the overall safety of personnel, the building, and environs. The consequences of credible incidents were considered and are discussed.

  11. Reliability analysis framework for computer-assisted medical decision systems

    International Nuclear Information System (INIS)

    Habas, Piotr A.; Zurada, Jacek M.; Elmaghraby, Adel S.; Tourassi, Georgia D.

    2007-01-01

    We present a technique that enhances computer-assisted decision (CAD) systems with the ability to assess the reliability of each individual decision they make. Reliability assessment is achieved by measuring the accuracy of a CAD system with known cases similar to the one in question. The proposed technique analyzes the feature space neighborhood of the query case to dynamically select an input-dependent set of known cases relevant to the query. This set is used to assess the local (query-specific) accuracy of the CAD system. The estimated local accuracy is utilized as a reliability measure of the CAD response to the query case. The underlying hypothesis of the study is that CAD decisions with higher reliability are more accurate. The above hypothesis was tested using a mammographic database of 1337 regions of interest (ROIs) with biopsy-proven ground truth (681 with masses, 656 with normal parenchyma). Three types of decision models, (i) a back-propagation neural network (BPNN), (ii) a generalized regression neural network (GRNN), and (iii) a support vector machine (SVM), were developed to detect masses based on eight morphological features automatically extracted from each ROI. The performance of all decision models was evaluated using the Receiver Operating Characteristic (ROC) analysis. The study showed that the proposed reliability measure is a strong predictor of the CAD system's case-specific accuracy. Specifically, the ROC area index for CAD predictions with high reliability was significantly better than for those with low reliability values. This result was consistent across all decision models investigated in the study. The proposed case-specific reliability analysis technique could be used to alert the CAD user when an opinion that is unlikely to be reliable is offered. The technique can be easily deployed in the clinical environment because it is applicable with a wide range of classifiers regardless of their structure and it requires neither additional

  12. Assessment of major nuclear technologies with decision and risk analysis

    International Nuclear Information System (INIS)

    Winterfeldt, D. von

    1995-01-01

    Selecting technologies for major nuclear programs involves several complexities, including multiple stakeholders, multiple conflicting objectives, uncertainties, and risk. In addition, the programmatic risks related to the schedule, cost, and performance of these technologies often become major issues in the selection process. This paper describes a decision analysis approach for addressing these complexities in a logical manner

  13. Multi-criteria decision analysis integrated with GIS for radio ...

    African Journals Online (AJOL)

    Multi-criteria decision analysis integrated with GIS for radio astronomical observatory site selection in peninsular of Malaysia. R Umar, Z.Z. Abidin, Z.A. Ibrahim, M.K.A. Kamarudin, S.N. Hazmin, A Endut, H Juahir ...

  14. Decision analysis and rational countermeasures in radiation protection

    International Nuclear Information System (INIS)

    Sinkko, K.

    1991-09-01

    During the past few years several international organizations (ICRP, IAEA, OECD/NEA), in revising their radiation protection principles, have emphasized the importance of the rationalization and planning of intervention after a nuclear accident. An accident itself and the introduction of protective action entails risks to the people affected, monetary costs and social disruption. Thus protective actions, often including objectives which are difficult to control simultaneously, cannot be undertaken without careful contemplation and consideration of the essential consequences of decisions. Often during an accident there is not enough time for careful consideration. Decision analysis is an analyzing and thought guiding method for the definition of objectives and comparison of options. It is an appropriate methodology assisting in rendering explicit and apparent all factors involved and evaluating their relative importance. The planning of intervention with the help of decision analysis is portion of the preparation for accident situations. In this report one of the techniques of decision analysis, multi-attribute utility analysis, is presented, as concerns its application in planning protective actions in the event of radiation accidents. (orig.)

  15. Decision analysis of Hanford underground storage tank waste retrieval systems

    International Nuclear Information System (INIS)

    Merkhofer, M.W.; Bitz, D.A.; Berry, D.L.; Jardine, L.J.

    1994-05-01

    A decision analysis approach has been proposed for planning the retrieval of hazardous, radioactive, and mixed wastes from underground storage tanks. This paper describes the proposed approach and illustrates its application to the single-shell storage tanks (SSTs) at Hanford, Washington

  16. A stochastic multicriteria model for evidence-based decision making in drug benefit-risk analysis.

    Science.gov (United States)

    Tervonen, Tommi; van Valkenhoef, Gert; Buskens, Erik; Hillege, Hans L; Postmus, Douwe

    2011-05-30

    Drug benefit-risk (BR) analysis is based on firm clinical evidence regarding various safety and efficacy outcomes. In this paper, we propose a new and more formal approach for constructing a supporting multi-criteria model that fully takes into account the evidence on efficacy and adverse drug reactions. Our approach is based on the stochastic multi-criteria acceptability analysis methodology, which allows us to compute the typical value judgments that support a decision, to quantify decision uncertainty, and to compute a comprehensive BR profile. We construct a multi-criteria model for the therapeutic group of second-generation antidepressants. We assess fluoxetine and venlafaxine together with placebo according to incidence of treatment response and three common adverse drug reactions by using data from a published study. Our model shows that there are clear trade-offs among the treatment alternatives. Copyright © 2011 John Wiley & Sons, Ltd.

  17. Procurement strategic analysis of nuclear safety equipment

    International Nuclear Information System (INIS)

    Wu Caixia; Yang Haifeng; Li Xiaoyang; Li Shixin

    2013-01-01

    The nuclear power development plan in China puts forward a challenge on procurement of nuclear safety equipment. Based on the characteristics of the procurement of nuclear safety equipment, requirements are raised for procurement process, including further clarification of equipment technical specification, establishment and improvement of the expert database of the nuclear power industry, adoption of more reasonable evaluation method and establishment of a unified platform for nuclear power plants to procure nuclear safety equipment. This paper makes recommendation of procurement strategy for nuclear power production enterprises from following aspects, making a plan of procurement progress, dividing procurement packages rationally, establishing supplier database through qualification review and implementing classified management, promoting localization process of key equipment continually and further improving the system and mechanism of procurement of nuclear safety equipment. (authors)

  18. The potential for meta-analysis to support decision analysis in ecology.

    Science.gov (United States)

    Mengersen, Kerrie; MacNeil, M Aaron; Caley, M Julian

    2015-06-01

    Meta-analysis and decision analysis are underpinned by well-developed methods that are commonly applied to a variety of problems and disciplines. While these two fields have been closely linked in some disciplines such as medicine, comparatively little attention has been paid to the potential benefits of linking them in ecology, despite reasonable expectations that benefits would be derived from doing so. Meta-analysis combines information from multiple studies to provide more accurate parameter estimates and to reduce the uncertainty surrounding them. Decision analysis involves selecting among alternative choices using statistical information that helps to shed light on the uncertainties involved. By linking meta-analysis to decision analysis, improved decisions can be made, with quantification of the costs and benefits of alternate decisions supported by a greater density of information. Here, we briefly review concepts of both meta-analysis and decision analysis, illustrating the natural linkage between them and the benefits from explicitly linking one to the other. We discuss some examples in which this linkage has been exploited in the medical arena and how improvements in precision and reduction of structural uncertainty inherent in a meta-analysis can provide substantive improvements to decision analysis outcomes by reducing uncertainty in expected loss and maximising information from across studies. We then argue that these significant benefits could be translated to ecology, in particular to the problem of making optimal ecological decisions in the face of uncertainty. Copyright © 2013 John Wiley & Sons, Ltd.

  19. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

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

    1995-04-01

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

  20. Probabilistic safety analysis and interpretation thereof

    International Nuclear Information System (INIS)

    Steininger, U.; Sacher, H.

    1999-01-01

    Increasing use of the instrumentation of PSA is being made in Germany for quantitative technical safety assessment, for example with regard to incidents which must be reported and forwarding of information, especially in the case of modification of nuclear plants. The Commission for Nuclear Reactor Safety recommends regular execution of PSA on a cycle period of ten years. According to the PSA guidance instructions, probabilistic analyses serve for assessing the degree of safety of the entire plant, expressed as the expectation value for the frequency of endangering conditions. The authors describe the method, action sequence and evaluation of the probabilistic safety analyses. The limits of probabilistic safety analyses arise in the practical implementation. Normally the guidance instructions for PSA are confined to the safety systems, so that in practice they are at best suitable for operational optimisation only to a limited extent. The present restriction of the analyses has a similar effect on power output operation of the plant. This seriously degrades the utilitarian value of these analyses for the plant operators. In order to further develop PSA as a supervisory and operational optimisation instrument, both authors consider it to be appropriate to bring together the specific know-how of analysts, manufacturers, plant operators and experts. (orig.) [de

  1. Risk Analysis and Decision Making FY 2013 Milestone Report

    Energy Technology Data Exchange (ETDEWEB)

    Engel, David W.; Dalton, Angela C.; Dale, Crystal; Jones, Edward; Thompson, J.

    2013-06-01

    Risk analysis and decision making is one of the critical objectives of CCSI, which seeks to use information from science-based models with quantified uncertainty to inform decision makers who are making large capital investments. The goal of this task is to develop tools and capabilities to facilitate the development of risk models tailored for carbon capture technologies, quantify the uncertainty of model predictions, and estimate the technical and financial risks associated with the system. This effort aims to reduce costs by identifying smarter demonstrations, which could accelerate development and deployment of the technology by several years.

  2. Accommodating complexity and human behaviors in decision analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Backus, George A.; Siirola, John Daniel; Schoenwald, David Alan; Strip, David R.; Hirsch, Gary B.; Bastian, Mark S.; Braithwaite, Karl R.; Homer, Jack [Homer Consulting

    2007-11-01

    This is the final report for a LDRD effort to address human behavior in decision support systems. One sister LDRD effort reports the extension of this work to include actual human choices and additional simulation analyses. Another provides the background for this effort and the programmatic directions for future work. This specific effort considered the feasibility of five aspects of model development required for analysis viability. To avoid the use of classified information, healthcare decisions and the system embedding them became the illustrative example for assessment.

  3. Systematic approaches to data analysis from the Critical Decision Method

    Directory of Open Access Journals (Sweden)

    Martin Sedlár

    2015-01-01

    Full Text Available The aim of the present paper is to introduce how to analyse the qualitative data from the Critical Decision Method. At first, characterizing the method provides the meaningful introduction into the issue. This method used in naturalistic decision making research is one of the cognitive task analysis methods, it is based on the retrospective semistructured interview about critical incident from the work and it may be applied in various domains such as emergency services, military, transport, sport or industry. Researchers can make two types of methodological adaptation. Within-method adaptations modify the way of conducting the interviews and cross-method adaptations combine this method with other related methods. There are many decsriptions of conducting the interview, but the descriptions how the data should be analysed are rare. Some researchers use conventional approaches like content analysis, grounded theory or individual procedures with reference to the objectives of research project. Wong (2004 describes two approaches to data analysis proposed for this method of data collection, which are described and reviewed in the details. They enable systematic work with a large amount of data. The structured approach organizes the data according to an a priori analysis framework and it is suitable for clearly defined object of research. Each incident is studied separately. At first, the decision chart showing the main decision points and then the incident summary are made. These decision points are used to identify the relevant statements from the transcript, which are analysed in terms of the Recognition-Primed Decision Model. Finally, the results from all the analysed incidents are integrated. The limitation of the structured approach is it may not reveal some interesting concepts. The emergent themes approach helps to identify these concepts while maintaining a systematic framework for analysis and it is used for exploratory research design. It

  4. Gap Analysis Approach for Construction Safety Program Improvement

    Directory of Open Access Journals (Sweden)

    Thanet Aksorn

    2007-06-01

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

  5. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

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

  6. 2005 dossier: granite. Tome: safety analysis of the geologic disposal

    International Nuclear Information System (INIS)

    2005-01-01

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

  7. Safety Lead Optimization and Candidate Identification: Integrating New Technologies into Decision-Making.

    Science.gov (United States)

    Dambach, Donna M; Misner, Dinah; Brock, Mathew; Fullerton, Aaron; Proctor, William; Maher, Jonathan; Lee, Dong; Ford, Kevin; Diaz, Dolores

    2016-04-18

    Discovery toxicology focuses on the identification of the most promising drug candidates through the development and implementation of lead optimization strategies and hypothesis-driven investigation of issues that enable rational and informed decision-making. The major goals are to [a] identify and progress the drug candidate with the best overall drug safety profile for a therapeutic area, [b] remove the most toxic drugs from the portfolio prior to entry into humans to reduce clinical attrition due to toxicity, and [c] establish a well-characterized hazard and translational risk profile to enable clinical trial designs. This is accomplished through a framework that balances the multiple considerations to identify a drug candidate with the overall best drug characteristics and provides a cogent understanding of mechanisms of toxicity. The framework components include establishing a target candidate profile for each program that defines the qualities of a successful candidate based on the intended therapeutic area, including the risk tolerance for liabilities; evaluating potential liabilities that may result from engaging the therapeutic target (pharmacology-mediated or on-target) and that are chemical structure-mediated (off-target); and characterizing identified liabilities. Lead optimization and investigation relies upon the integrated use of a variety of technologies and models (in silico, in vitro, and in vivo) that have achieved a sufficient level of qualification or validation to provide confidence in their use. We describe the strategic applications of various nonclinical models (established and new) for a holistic and integrated risk assessment that is used for rational decision-making. While this review focuses on strategies for small molecules, the overall concepts, approaches, and technologies are generally applicable to biotherapeutics.

  8. A framework for sensitivity analysis of decision trees.

    Science.gov (United States)

    Kamiński, Bogumił; Jakubczyk, Michał; Szufel, Przemysław

    2018-01-01

    In the paper, we consider sequential decision problems with uncertainty, represented as decision trees. Sensitivity analysis is always a crucial element of decision making and in decision trees it often focuses on probabilities. In the stochastic model considered, the user often has only limited information about the true values of probabilities. We develop a framework for performing sensitivity analysis of optimal strategies accounting for this distributional uncertainty. We design this robust optimization approach in an intuitive and not overly technical way, to make it simple to apply in daily managerial practice. The proposed framework allows for (1) analysis of the stability of the expected-value-maximizing strategy and (2) identification of strategies which are robust with respect to pessimistic/optimistic/mode-favoring perturbations of probabilities. We verify the properties of our approach in two cases: (a) probabilities in a tree are the primitives of the model and can be modified independently; (b) probabilities in a tree reflect some underlying, structural probabilities, and are interrelated. We provide a free software tool implementing the methods described.

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

    Science.gov (United States)

    Lyons, Vanessa E; Popejoy, Lori L

    2014-02-01

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

  10. Ethical analysis to improve decision-making on health technologies

    DEFF Research Database (Denmark)

    Saarni, Samuli I; Hofmann, Bjørn; Lampe, Kristian

    2008-01-01

    Health technology assessment (HTA) is the multidisciplinary study of the implications of the development, diffusion and use of health technologies. It supports health-policy decisions by providing a joint knowledge base for decision-makers. To increase its policy relevance, HTA tries to extend...... beyond effectiveness and costs to also considering the social, organizational and ethical implications of technologies. However, a commonly accepted method for analysing the ethical aspects of health technologies is lacking. This paper describes a model for ethical analysis of health technology...... to only analyse the ethical consequences of a technology, but also the ethical issues of the whole HTA process must be considered. Selection of assessment topics, methods and outcomes is essentially a value-laden decision. Health technologies may challenge moral or cultural values and beliefs...

  11. A dynamic probabilistic safety margin characterization approach in support of Integrated Deterministic and Probabilistic Safety Analysis

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Rai, Ajit; Zio, Enrico

    2016-01-01

    The challenge of Risk-Informed Safety Margin Characterization (RISMC) is to develop a methodology for estimating system safety margins in the presence of stochastic and epistemic uncertainties affecting the system dynamic behavior. This is useful to support decision-making for licensing purposes. In the present work, safety margin uncertainties are handled by Order Statistics (OS) (with both Bracketing and Coverage approaches) to jointly estimate percentiles of the distributions of the safety parameter and of the time required for it to reach these percentiles values during its dynamic evolution. The novelty of the proposed approach consists in the integration of dynamic aspects (i.e., timing of events) into the definition of a dynamic safety margin for a probabilistic Quantification of Margin and Uncertainties (QMU). The system here considered for demonstration purposes is the Lead–Bismuth Eutectic- eXperimental Accelerator Driven System (LBE-XADS). - Highlights: • We integrate dynamic aspects into the definition of a safety margins. • We consider stochastic and epistemic uncertainties affecting the system dynamics. • Uncertainties are handled by Order Statistics (OS). • We estimate the system grace time during accidental scenarios. • We apply the approach to an LBE-XADS accidental scenario.

  12. First wall and blanket module safety enhancement by material selection and design decision

    International Nuclear Information System (INIS)

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems

  13. Using the Situated Clinical Decision-Making framework to guide analysis of nurses' clinical decision-making.

    Science.gov (United States)

    Gillespie, Mary

    2010-11-01

    Nurses' clinical decision-making is a complex process that holds potential to influence the quality of care provided and patient outcomes. The evolution of nurses' decision-making that occurs with experience has been well documented. In addition, literature includes numerous strategies and approaches purported to support development of nurses' clinical decision-making. There has been, however, significantly less attention given to the process of assessing nurses' clinical decision-making and novice clinical educators are often challenged with knowing how to best support nurses and nursing students in developing their clinical decision-making capacity. The Situated Clinical Decision-Making framework is presented for use by clinical educators: it provides a structured approach to analyzing nursing students' and novice nurses' decision-making in clinical nursing practice, assists educators in identifying specific issues within nurses' clinical decision-making, and guides selection of relevant strategies to support development of clinical decision-making. A series of questions is offered as a guide for clinical educators when assessing nurses' clinical decision-making. The discussion presents key considerations related to analysis of various decision-making components, including common sources of challenge and errors that may occur within nurses' clinical decision-making. An exemplar illustrates use of the framework and guiding questions. Implications of this approach for selection of strategies that support development of clinical decision-making are highlighted. Copyright © 2010 Elsevier Ltd. All rights reserved.

  14. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    BERGMANN, D.W.

    1999-01-01

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  15. Probabilistic safety analysis : a new nuclear power plants licensing method

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de.

    1982-04-01

    After a brief retrospect of the application of Probabilistic Safety Analysis in the nuclear field, the basic differences between the deterministic licensing method, currently in use, and the probabilistic method are explained. Next, the two main proposals (by the AIF and the ACRS) concerning the establishment of the so-called quantitative safety goals (or simply 'safety goals') are separately presented and afterwards compared in their most fundamental aspects. Finally, some recent applications and future possibilities are discussed. (Author) [pt

  16. Safety analysis report for the Waste Storage Facility. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  17. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  18. Multivariate analysis of flow cytometric data using decision trees.

    Science.gov (United States)

    Simon, Svenja; Guthke, Reinhard; Kamradt, Thomas; Frey, Oliver

    2012-01-01

    Characterization of the response of the host immune system is important in understanding the bidirectional interactions between the host and microbial pathogens. For research on the host site, flow cytometry has become one of the major tools in immunology. Advances in technology and reagents allow now the simultaneous assessment of multiple markers on a single cell level generating multidimensional data sets that require multivariate statistical analysis. We explored the explanatory power of the supervised machine learning method called "induction of decision trees" in flow cytometric data. In order to examine whether the production of a certain cytokine is depended on other cytokines, datasets from intracellular staining for six cytokines with complex patterns of co-expression were analyzed by induction of decision trees. After weighting the data according to their class probabilities, we created a total of 13,392 different decision trees for each given cytokine with different parameter settings. For a more realistic estimation of the decision trees' quality, we used stratified fivefold cross validation and chose the "best" tree according to a combination of different quality criteria. While some of the decision trees reflected previously known co-expression patterns, we found that the expression of some cytokines was not only dependent on the co-expression of others per se, but was also dependent on the intensity of expression. Thus, for the first time we successfully used induction of decision trees for the analysis of high dimensional flow cytometric data and demonstrated the feasibility of this method to reveal structural patterns in such data sets.

  19. Decision Making in Nursing Practice: A Concept Analysis.

    Science.gov (United States)

    Johansen, Mary L; O'Brien, Janice L

    2016-01-01

    The study aims to gain an understanding of the concept of decision making as it relates to the nurse practice environment. Rodgers' evolutionary method on concept analysis was used as a framework for the study of the concept. Articles from 1952 to 2014 were reviewed from PsycINFO, Medline, Cumulative Index to Nursing and Allied Health Literature (CINAHL), JSTOR, PubMed, and Science Direct. Findings suggest that decision making in the nurse practice environment is a complex process, integral to the nursing profession. The definition of decision making, and the attributes, antecedents, and consequences, are discussed. Contextual factors that influence the process are also discussed. An exemplar is presented to illustrate the concept. Decision making in the nurse practice environment is a dynamic conceptual process that may affect patient outcomes. Nurses need to call upon ways of knowing to make sound decisions and should be self-reflective in order to develop the process further in the professional arena. The need for further research is discussed. © 2015 Wiley Periodicals, Inc.

  20. Applying importance-performance analysis to patient safety culture.

    Science.gov (United States)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Rohar, S.

    1999-01-01

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

  2. Risk regulation in environment, health and safety : Decision in the face of uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Ettlinger, L A [The Oxford Group, Baltimore, MD (United States)

    1999-12-01

    unlikely to be resolved solely by administrative reform, reliance on experts, or risk education because, generally, these reforms are focused strictly on narrowing the risk uncertainty. From economic principles, however, we recognize that the value of efforts to minimize this risk uncertainty (or to maximize the technical and scientific information about risks) is subject to 'diminishing marginal utility'. Thus, these efforts ultimately can be expected to become unpopular with both citizens and elected officials. We expect, therefore, that the outcome of the risk regulation debate, and ultimately environment, health and safety policy decisions, will primarily be determined by the values of the proponents and opponents as to the relative ranking of efficiency and equity. Most likely, the best means available for reforming risk regulation and making it more coherent and to avoid some of the costly mistakes of the past is to: 1. increase the resources available for identifying risks and effective responses to them, 2. build in mechanisms for informing the public that policy decisions using these regulations are value laden, and 3. publicize both the process and the outcomes of risk regulations and policy decisions.

  3. Risk regulation in environment, health and safety : Decision in the face of uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Ettlinger, L.A. [The Oxford Group, Baltimore, MD (United States)

    1999-12-01

    relationship are unlikely to be resolved solely by administrative reform, reliance on experts, or risk education because, generally, these reforms are focused strictly on narrowing the risk uncertainty. From economic principles, however, we recognize that the value of efforts to minimize this risk uncertainty (or to maximize the technical and scientific information about risks) is subject to 'diminishing marginal utility'. Thus, these efforts ultimately can be expected to become unpopular with both citizens and elected officials. We expect, therefore, that the outcome of the risk regulation debate, and ultimately environment, health and safety policy decisions, will primarily be determined by the values of the proponents and opponents as to the relative ranking of efficiency and equity. Most likely, the best means available for reforming risk regulation and making it more coherent and to avoid some of the costly mistakes of the past is to: 1. increase the resources available for identifying risks and effective responses to them, 2. build in mechanisms for informing the public that policy decisions using these regulations are value laden, and 3. publicize both the process and the outcomes of risk regulations and policy decisions.

  4. Risk regulation in environment, health and safety : Decision in the face of uncertainty

    International Nuclear Information System (INIS)

    Ettlinger, L.A.

    1999-01-01

    unlikely to be resolved solely by administrative reform, reliance on experts, or risk education because, generally, these reforms are focused strictly on narrowing the risk uncertainty. From economic principles, however, we recognize that the value of efforts to minimize this risk uncertainty (or to maximize the technical and scientific information about risks) is subject to 'diminishing marginal utility'. Thus, these efforts ultimately can be expected to become unpopular with both citizens and elected officials. We expect, therefore, that the outcome of the risk regulation debate, and ultimately environment, health and safety policy decisions, will primarily be determined by the values of the proponents and opponents as to the relative ranking of efficiency and equity. Most likely, the best means available for reforming risk regulation and making it more coherent and to avoid some of the costly mistakes of the past is to: 1. increase the resources available for identifying risks and effective responses to them, 2. build in mechanisms for informing the public that policy decisions using these regulations are value laden, and 3. publicize both the process and the outcomes of risk regulations and policy decisions

  5. An analysis of safety control effectiveness

    International Nuclear Information System (INIS)

    Son, K.S.; Melchers, R.E.; Kal, W.M.

    2000-01-01

    The cost of injuries and 'accidents' to an organisation is very important in establishing how much it should spend on safety control. Despite the usefulness of information about the cost of a company's accidents, it is not customary accounting practice to make these data available. Of the two kinds of costs incurred by a company through occupational injuries and accidents, direct costs and indirect costs; the direct costs are much easier to estimate. However, the uninsured costs are usually more critical and should be estimated by each company. The authors investigate a general model to estimate the above costs and hence to establish efficient safety control. One construction company has been a pilot for this study. By analysing actual company data for three years, it is found that the efficient safety control cost should be 1.2-1.3% of total contract costs

  6. SEDIMENT ANALYSIS NETWORK FOR DECISION SUPPORT (SANDS) LANDSAT GEOLOGICAL SURVEY OF AL (GSA) ANALYSIS V1

    Data.gov (United States)

    National Aeronautics and Space Administration — The Sediment Analysis Network for Decision Support (SANDS) Landsat Geological Survey of AL (GSA) Analysis dataset analyzed changes in the coastal shoreline and...

  7. Job safety and awareness analysis of safety implementation among electrical workers in airport service company

    Directory of Open Access Journals (Sweden)

    Putra Perdana Suteja

    2018-01-01

    Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

  8. Multiple attribute decision making model and application to food safety risk evaluation.

    Science.gov (United States)

    Ma, Lihua; Chen, Hong; Yan, Huizhe; Yang, Lifeng; Wu, Lifeng

    2017-01-01

    Decision making for supermarket food purchase decisions are characterized by network relationships. This paper analyzed factors that influence supermarket food selection and proposes a supplier evaluation index system based on the whole process of food production. The author established the intuitive interval value fuzzy set evaluation model based on characteristics of the network relationship among decision makers, and validated for a multiple attribute decision making case study. Thus, the proposed model provides a reliable, accurate method for multiple attribute decision making.

  9. Safety analysis of the UTSI-CFFF superconducting magnet

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Galileo and Ulysses missions safety analysis and launch readiness status

    International Nuclear Information System (INIS)

    Cork, M.J.; Turi, J.A.

    1989-01-01

    The Galileo spacecraft will explore the Jupiter system and Ulysses will fly by Jupiter en route to a polar orbit of the sun. Both spacecraft are powered by general purpose heat source radioisotope thermoelectric generators (RTGs). As a result of the Challenger accident and subsequent mission reprogramming, the Galileo and Ulysses missions' safety analysis had to be repeated. In addition to presenting an overview of the safety analysis status for the missions, this paper presents a brief review of the missions' objectives and design approaches, RTG design characteristics and development history, and a description of the safety analysis process. (author)

  11. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

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

  12. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  13. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo; Seong, Poong Hyun

    1997-01-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formed safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system

  14. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

    Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

  15. Safety analysis of passing maneuvers using extreme value theory

    Directory of Open Access Journals (Sweden)

    Haneen Farah

    2017-04-01

    The results indicate that this is a promising approach for safety evaluation. On-going work of the authors will attempt to generalize this method to other safety measures related to passing maneuvers, test it for the detailed analysis of the effect of demographic factors on passing maneuvers' crash probability and for its usefulness in a traffic simulation environment.

  16. Integrated program of using of Probabilistic Safety Analysis in Spain

    International Nuclear Information System (INIS)

    1998-01-01

    Since 25 June 1986, when the CSN (Nuclear Safety Conseil) approve the Integrated Program of Probabilistic Safety Analysis, this program has articulated the main activities of CSN. This document summarize the activities developed during these years and reviews the Integrated programme

  17. Systems Analysis of NASA Aviation Safety Program: Final Report

    Science.gov (United States)

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

    2013-01-01

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

  18. OASIS: An automotive analysis and safety engineering instrument

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  19. Design of Graph Analysis Model to support Decision Making

    International Nuclear Information System (INIS)

    An, Sang Ha; Lee, Sung Jin; Chang, Soon Heung; Kim, Sung Ho; Kim, Tae Woon

    2005-01-01

    Korea is meeting the growing electric power needs by using nuclear, fissile, hydro energy and so on. But we can not use fissile energy forever, and the people's consideration about nature has been changed. So we have to prepare appropriate energy by the conditions before people need more energy. And we should prepare dynamic response because people's need would be changed as the time goes on. So we designed graphic analysis model (GAM) for the dynamic analysis of decision on the energy sources. It can support Analytic Hierarchy Process (AHP) analysis based on Graphic User Interface

  20. Human reliability analysis in Loviisa probabilistic safety analysis

    International Nuclear Information System (INIS)

    Illman, L.; Isaksson, J.; Makkonen, L.; Vaurio, J.K.; Vuorio, U.

    1986-01-01

    The human reliability analysis in the Loviisa PSA project is carried out for three major groups of errors in human actions: (A) errors made before an initiating event, (B) errors that initiate a transient and (C) errors made during transients. Recovery possibilities are also included in each group. The methods used or planned for each group are described. A simplified THERP approach is used for group A, with emphasis on test and maintenance error recovery aspects and dependencies between redundancies. For group B, task analyses and human factors assessments are made for startup, shutdown and operational transients, with emphasis on potential common cause initiators. For group C, both misdiagnosis and slow decision making are analyzed, as well as errors made in carrying out necessary or backup actions. New or advanced features of the methodology are described

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

    International Nuclear Information System (INIS)

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  3. An analysis of the traffic safety phenomenon.

    NARCIS (Netherlands)

    Asmussen, E. & Kranenburg, A.

    1982-01-01

    The lack of traffic safety is a combination of the critical coincidence of circumstances in the traffic of incidents (near-accidents) and accidents with unwanted (permanent) consequences, such as fatalities, injured and disabled persons and material damage. This definition covers the whole of the

  4. Safety Analysis of Stochastic Dynamical Systems

    DEFF Research Database (Denmark)

    Sloth, Christoffer; Wisniewski, Rafael

    2015-01-01

    This paper presents a method for verifying the safety of a stochastic system. In particular, we show how to compute the largest set of initial conditions such that a given stochastic system is safe with probability p. To compute the set of initial conditions we rely on the moment method that via...... that shows how the p-safe initial set is computed numerically....

  5. Preliminary safety design analysis of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    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 document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

  6. Analysis of high-pressure safety valves

    NARCIS (Netherlands)

    Beune, A.

    2009-01-01

    In presently used safety valve sizing standards the gas discharge capacity is based on a nozzle flow derived from ideal gas theory. At high pressures or low temperatures real gas effects can no longer be neglected, so the discharge coefficient corrected for flow losses cannot be assumed constant

  7. HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

    International Nuclear Information System (INIS)

    EVANS, C.B.

    2004-01-01

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

  8. Operational safety analysis status of Novi Han repository

    International Nuclear Information System (INIS)

    Boiadjiev, A.

    2000-01-01

    This article presents the status of the safety studies and activities related to Novi Han repository. The case of this facility is such that no clear boundary exists between post-closure safety assessment and operational safety assessment. The major findings of these activities are given. The Safety Analysis Report (SAR) for Novi Han repository is developed by Risk Engineering Ltd. under a contract with the Committee on the Use of Atomic Energy for Peaceful Purposes. The general structure and main conclusions and recommendations of the SAR are presented. (author)

  9. Analysis for making a regulatory decision to equipment of industrial gammagraphy in Argentin; Analisis para la toma de decision regulatoria sobre equipos de gammagrafia industrial en Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Ermacora, Marcela G.; Vidal, Dora N.; Alonso, Maria T., E-mail: mermacora@arn.gob.ar, E-mail: dvidal@arn.gob.ar, E-mail: malonso@arn.gob.ar [Autoridad Regulatoria Nuclear (ARN), Buenos Aires (Argentina)

    2013-10-01

    Industrial gammagraphy is a practice widely used as a nondestructive testing technique in Argentina. Experience worldwide has shown the need for an improvement in the intrinsic safety of the equipment used in this lab. In response to this reason, the board of the Nuclear Regulatory Authority (ARN) has considered a proposal to withdraw service movement and much of the equipment inventory scan belonging to industrial facilities nationwide. The main objective of this paper is to present the results of the analysis performed to support the above proposal. The main elements of evaluation can be summarized as follows: I) the teams that do not conform to international recommendations regarding compliance with key safety requirements of international standards such as ISO 3999:2004 (E) {sup R}adiation protection - Industrial Apparatus for gamma radiography - Specifications for performance, design and tests {sup ;} II) the decision by some manufacturers to discontinue production of certain models of equipment and the provision of spare parts, and III) the validity of certificates bulk type B (U) for transport. In conclusion, it highlights the importance of a regulatory decision supplementary to the Standard AR 7.9.1 concerning the operation of scan equipment industry, based on current international recommendations and Argentina's commitment to good practice and safety culture which can lead to a positive impact on radiation safety in this art.

  10. Using multicriteria decision analysis during drug development to predict reimbursement decisions.

    Science.gov (United States)

    Williams, Paul; Mauskopf, Josephine; Lebiecki, Jake; Kilburg, Anne

    2014-01-01

    Pharmaceutical companies design clinical development programs to generate the data that they believe will support reimbursement for the experimental compound. The objective of the study was to present a process for using multicriteria decision analysis (MCDA) by a pharmaceutical company to estimate the probability of a positive recommendation for reimbursement for a new drug given drug and environmental attributes. The MCDA process included 1) selection of decisions makers who were representative of those making reimbursement decisions in a specific country; 2) two pre-workshop questionnaires to identify the most important attributes and their relative importance for a positive recommendation for a new drug; 3) a 1-day workshop during which participants undertook three tasks: i) they agreed on a final list of decision attributes and their importance weights, ii) they developed level descriptions for these attributes and mapped each attribute level to a value function, and iii) they developed profiles for hypothetical products 'just likely to be reimbursed'; and 4) use of the data from the workshop to develop a prediction algorithm based on a logistic regression analysis. The MCDA process is illustrated using case studies for three countries, the United Kingdom, Germany, and Spain. The extent to which the prediction algorithms for each country captured the decision processes for the workshop participants in our case studies was tested using a post-meeting questionnaire that asked the participants to make recommendations for a set of hypothetical products. The data collected in the case study workshops resulted in a prediction algorithm: 1) for the United Kingdom, the probability of a positive recommendation for different ranges of cost-effectiveness ratios; 2) for Spain, the probability of a positive recommendation at the national and regional levels; and 3) for Germany, the probability of a determination of clinical benefit. The results from the post

  11. Revealed preferences towards the appraisal of orphan drugs in Poland - multi criteria decision analysis.

    Science.gov (United States)

    Kolasa, Katarzyna; Zwolinski, Krzysztof Miroslaw; Zah, Vladimir; Kaló, Zoltán; Lewandowski, Tadeusz

    2018-04-27

    A Multi Criteria Decision Analysis (MCDA) technique was adopted to reveal the preferences of the Appraisal Body of the Polish HTA agency towards orphan drugs (OMPs). There were 34 positive and 23 negative HTA recommendations out of 54 distinctive drug-indication pairs. The MCDA matrix consisted of 13 criteria, seven of which made the most impact on the HTA process. Appraisal of clinical evidence, cost of therapy, and safety considerations were the main contributors to the HTA guidance, whilst advancement of technology and manufacturing costs made the least impact. MCDA can be regarded as a valuable tool for revealing decision makers' preferences in the healthcare sector. Given that only roughly half of all criteria included in the MCDA matrix were deemed to make an impact on the HTA process, there is certainly some room for improvement with respect to the adaptation of a new approach towards the value assessment of OMPs in Poland.

  12. Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework

    International Nuclear Information System (INIS)

    Cappelli, M.; Gadomski, A. M.; Sepiellis, M.; Wronikowska, M. W.

    2012-01-01

    In the field of nuclear power plant (NPP) safety modeling, the perception of the role of socio-cognitive engineering (SCE) is continuously increasing. Today, the focus is especially on the identification of human and organization decisional errors caused by operators and managers under high-risk conditions, as evident by analyzing reports on nuclear incidents occurred in the past. At present, the engineering and social safety requirements need to enlarge their domain of interest in such a way to include all possible losses generating events that could be the consequences of an abnormal state of a NPP. Socio-cognitive modeling of Integrated Nuclear Safety Management (INSM) using the TOGA meta-theory has been discussed during the ICCAP 2011 Conference. In this paper, more detailed aspects of the cognitive decision-making and its possible human errors and organizational vulnerability are presented. The formal TOGA-based network model for cognitive decision-making enables to indicate and analyze nodes and arcs in which plant operators and managers errors may appear. The TOGA's multi-level IPK (Information, Preferences, Knowledge) model of abstract intelligent agents (AIAs) is applied. In the NPP context, super-safety approach is also discussed, by taking under consideration unexpected events and managing them from a systemic perspective. As the nature of human errors depends on the specific properties of the decision-maker and the decisional context of operation, a classification of decision-making using IPK is suggested. Several types of initial situations of decision-making useful for the diagnosis of NPP operators and managers errors are considered. The developed models can be used as a basis for applications to NPP educational or engineering simulators to be used for training the NPP executive staff. (authors)

  13. Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework

    Energy Technology Data Exchange (ETDEWEB)

    Cappelli, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Gadomski, A. M. [ECONA, Centro Interuniversitario Elaborazione Cognitiva Sistemi Naturali e Artificiali, via dei Marsi 47, Rome (Italy); Sepiellis, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Wronikowska, M. W. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Poznan School of Social Sciences (Poland)

    2012-07-01

    In the field of nuclear power plant (NPP) safety modeling, the perception of the role of socio-cognitive engineering (SCE) is continuously increasing. Today, the focus is especially on the identification of human and organization decisional errors caused by operators and managers under high-risk conditions, as evident by analyzing reports on nuclear incidents occurred in the past. At present, the engineering and social safety requirements need to enlarge their domain of interest in such a way to include all possible losses generating events that could be the consequences of an abnormal state of a NPP. Socio-cognitive modeling of Integrated Nuclear Safety Management (INSM) using the TOGA meta-theory has been discussed during the ICCAP 2011 Conference. In this paper, more detailed aspects of the cognitive decision-making and its possible human errors and organizational vulnerability are presented. The formal TOGA-based network model for cognitive decision-making enables to indicate and analyze nodes and arcs in which plant operators and managers errors may appear. The TOGA's multi-level IPK (Information, Preferences, Knowledge) model of abstract intelligent agents (AIAs) is applied. In the NPP context, super-safety approach is also discussed, by taking under consideration unexpected events and managing them from a systemic perspective. As the nature of human errors depends on the specific properties of the decision-maker and the decisional context of operation, a classification of decision-making using IPK is suggested. Several types of initial situations of decision-making useful for the diagnosis of NPP operators and managers errors are considered. The developed models can be used as a basis for applications to NPP educational or engineering simulators to be used for training the NPP executive staff. (authors)

  14. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Multi-criteria decision analysis: Limitations, pitfalls, and practical difficulties

    Energy Technology Data Exchange (ETDEWEB)

    Kujawski, Edouard

    2003-02-01

    The 2002 Winter Olympics women's figure skating competition is used as a case study to illustrate some of the limitations, pitfalls, and practical difficulties of Multi-Criteria Decision Analysis (MCDA). The paper compares several widely used models for synthesizing the multiple attributes into a single aggregate value. The various MCDA models can provide conflicting rankings of the alternatives for a common set of information even under states of certainty. Analysts involved in MCDA need to deal with the following challenging tasks: (1) selecting an appropriate analysis method, and (2) properly interpreting the results. An additional trap is the availability of software tools that implement specific MCDA models that can beguile the user with quantitative scores. These conclusions are independent of the decision domain and they should help foster better MCDA practices in many fields including systems engineering trade studies.

  17. Cost-effectiveness analysis: what it really means for transfusion medicine decision making.

    Science.gov (United States)

    Custer, Brian; Hoch, Jeffrey S

    2009-01-01

    Some have suggested that "blood is different," and the role for cost-effectiveness is thus circumscribed. In this article, the authors start by reviewing key concepts in health economics and economic analysis methods. Examples are drawn from published blood safety studies. After explaining the underlying reasoning behind cost-effectiveness analysis, the authors point out how economic thinking is evident in some aspects of transfusion medicine. Some cost-effectiveness study results for blood safety are discussed to provide context, followed by consideration of prominent decisions that have been made in transfusion medicine field. In the last section, the authors conjecture as to why in some cases cost-effectiveness analysis appears to have greater impact than in others, noting the terrible price that is paid in mortality and morbidity when cost-effectiveness analysis is ignored. In this context, the implications of opportunity cost are discussed, and it is noted that opportunity cost should not be viewed as benefits forgone by concentrating on one aspect of blood safety and instead should be viewed as our societal willingness to misallocate resources to achieve less health for the same cost.

  18. Ship Engine Room Casualty Analysis by Using Decision Tree Method

    Directory of Open Access Journals (Sweden)

    Ömür Yaşar SAATÇİOĞLU

    2017-03-01

    Full Text Available Ships may encounter undesirable conditions during operations. In consequence of a casualty, fire, explosion, flooding, grounding, injury even death may occur. Besides, these results can be avoidable with precautions and preventive operating processes. In maritime transportation, casualties depend on various factors. These were listed as misuse of the engine equipment and tools, defective machinery or equipment, inadequacy of operational procedure and measure of safety and force majeure effects. Casualty reports which were published in Australia, New Zealand, United Kingdom, Canada and United States until 2015 were examined and the probable causes and consequences of casualties were determined with their occurrence percentages. In this study, 89 marine investigation reports regarding engine room casualties were analyzed. Casualty factors were analyzed with their frequency percentages and also their main causes were constructed. This study aims to investigate engine room based casualties, frequency of each casualty type and main causes by using decision tree method.

  19. Municipal solid waste management system: decision support through systems analysis

    OpenAIRE

    Pires, Ana Lúcia Lourenço

    2010-01-01

    Thesis submitted to the Universidade Nova de Lisboa, Faculdade de Ciências e Tecnologia for the degree of Doctor of Philosophy in Environmental Engineering The present study intends to show the development of systems analysis model applied to solid waste management system, applied into AMARSUL, a solid waste management system responsible for the management of municipal solid waste produced in Setúbal peninsula, Portugal. The model developed intended to promote sustainable decision making, ...

  20. Renewable Energy Data, Analysis, and Decisions: A Guide for Practitioners

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Sarah L [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Lopez, Anthony J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Watson, Andrea C [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Grue, Nicholas W [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Leisch, Jennifer E. [United States Agency for International Development (USAID)

    2018-03-16

    High-quality renewable energy resource data and other geographic information system (GIS) data are essential for the transition to a clean energy economy that prioritizes local resources, improves resiliency, creates jobs, and promotes energy independence. This guide is intended to support policymakers and planners, as well as technical experts, consultants, and academics in incorporating improved data and analysis into renewable energy decision-making.