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.

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

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.)

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.

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

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

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.)

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)

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)

Regulatory decision making by decision analyses

International Nuclear Information System (INIS)

Holmberg, J.; Pulkkinen, U.

1993-11-01

The Technical Research Centre of Finland (VTT) has studied with the Finnish Centre for Radiation and Nuclear Safety (STUK) the applicability of decision analytic approach to the treatment of nuclear safety related problems at the regulatory body. The role of probabilistic safety assessment (PSA) in decision making has also been discussed. In the study, inspectors from STUK exercised with a decision analytic approach by reoperationalizing two occurred and solved problems. The research scientist from VTT acted as systems analysts guiding the analysis process. The first case was related to a common cause failure phenomenon in solenoid valves controlling pneumatic valves important to safety of the plant. The problem of the regulatory body was to judge whether to allow continued operation or to require more detailed inspections and in which time chedule the inspections should be done. The latter problem was to evaluate design changes of external electrical grid connections after a fire incident had revealed weakness in the separation of electrical system. In both cases, the decision analysis was carried out several sessions in which decision makers, technical experts as well as experts of decision analysis participated. A multi-attribute value function was applied as a decision model so that attributes had to be defined to quantify the levels of achievements of the objectives. The attributes included both indicators related to the level of operational safety of the plant such as core damage frequency given by PSA, and indicators related to the safety culture, i.e., how well the chosen option fits on the regulatory policy. (24 refs., 6 figs., 9 tabs.)

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)

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

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.

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

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

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.

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

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

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

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.

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

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.

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.)

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…

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)

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.

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

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

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

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

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.

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)

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

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

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.

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.

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

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.

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

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

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,

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

Analysis respons to the implementation of nuclear installations safety culture using AHP-TOPSIS

Science.gov (United States)

Situmorang, J.; Kuntoro, I.; Santoso, S.; Subekti, M.; Sunaryo, G. R.

2018-02-01

An analysis of responses to the implementation of nuclear installations safety culture has been done using AHP (Analitic Hierarchy Process) - TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). Safety culture is considered as collective commitments of the decision-making level, management level, and individual level. Thus each level will provide a subjective perspective as an alternative approach to implementation. Furthermore safety culture is considered by the statement of five characteristics which in more detail form consist of 37 attributes, and therefore can be expressed as multi-attribute state. Those characteristics and or attributes will be a criterion and its value is difficult to determine. Those criteria of course, will determine and strongly influence the implementation of the corresponding safety culture. To determine the pattern and magnitude of the influence is done by using a TOPSIS that is based on decision matrix approach and is composed of alternatives and criteria. The weight of each criterion is determined by AHP technique. The data used are data collected through questionnaires at the workshop on safety and health in 2015. .Reliability test of data gives Cronbah Alpha value of 95.5% which according to the criteria is stated reliable. Validity test using bivariate correlation analysis technique between each attribute give Pearson correlation for all attribute is significant at level 0,01. Using confirmatory factor analysis gives Kaise-Meyer-Olkin of sampling Adequacy (KMO) is 0.719 and it is greater than the acceptance criterion 0.5 as well as the 0.000 significance level much smaller than 0.05 and stated that further analysis could be performed. As a result of the analysis it is found that responses from the level of decision maker (second echelon) dominate the best order preference rank to be the best solution in strengthening the nuclear installation safety culture, except for the first characteristics, safety is a

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)

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.)

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

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)

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)

The use of case tools in OPG safety analysis code qualification

International Nuclear Information System (INIS)

Pascoe, J.; Cheung, A.; Westbye, C.

2001-01-01

Ontario Power Generation (OPG) is currently qualifying its critical safety analysis software. The software quality assurance (SQA) framework is described. Given the legacy nature of much of the safety analysis software the reverse engineering methodology has been adopted. The safety analysis suite of codes was developed over a period of many years to differing standards of quality and had sparse or incomplete documentation. Key elements of the reverse engineering process require recovery of design information from existing coding. This recovery, if performed manually, could represent an enormous effort. Driven by a need to maximize productivity and enhance the repeatability and objectivity of software qualification activities the decision was made to acquire or develop and implement Computer Aided Software Engineering (CASE) tools. This paper presents relevant background information on CASE tools and discusses how the OPG SQA requirements were used to assess the suitability of available CASE tools. Key findings from the application of CASE tools to the qualification of the OPG safety analysis software are discussed. (author)

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.

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

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.)

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

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.)

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

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

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

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

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.

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.

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.

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...

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.)

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.)

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.)

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

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

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.

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

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

Archetypes for Organisational Safety

Science.gov (United States)

Marais, Karen; Leveson, Nancy G.

2003-01-01

We propose a framework using system dynamics to model the dynamic behavior of organizations in accident analysis. Most current accident analysis techniques are event-based and do not adequately capture the dynamic complexity and non-linear interactions that characterize accidents in complex systems. In this paper we propose a set of system safety archetypes that model common safety culture flaws in organizations, i.e., the dynamic behaviour of organizations that often leads to accidents. As accident analysis and investigation tools, the archetypes can be used to develop dynamic models that describe the systemic and organizational factors contributing to the accident. The archetypes help clarify why safety-related decisions do not always result in the desired behavior, and how independent decisions in different parts of the organization can combine to impact safety.

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.

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

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

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

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

Corporate financial decision makers' perceptions of their company's safety performance, programs and personnel: Do company size and industry injury risk matter?

Science.gov (United States)

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

2010-01-01

Top-level managers make important decisions about safety-related issues, yet little research has been done involving these individuals. The current study explored corporate financial decisions makers' perceptions of their company's safety and their justifications for these perceptions. This study also explored whether their perceptions and justifications varied as a function of company size or industry injury risk. A total of 404 individuals who were the most senior managers responsible for making decisions about property and casualty risk at their companies participated in this study. The participants took part in a telephone survey. The results suggest that corporate financial decision makers have positive views of safety at their companies relative to safety at other companies within their industries. Further, many believe their company's safety is influenced by the attention/emphasis placed on safety and the selection and training of safety personnel. Participants' perceptions varied somewhat based on the size of their company and the level of injury risk in their industry. While definitive conclusions about corporate financial decision makers' perceptions of safety cannot be reached as a result of this single study, this work does lay groundwork for future research aimed at better understanding the perceptions top-level managers.

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)

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

Kim, Yun Goo; Oh, Eung Se

2016-01-01

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

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.

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.)

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)

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)

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.)

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)

The role of social cost-benefit analysis in societal decision-making under large uncertainties with application to robbery at a cash depot

International Nuclear Information System (INIS)

Jones-Lee, M.; Aven, T.

2009-01-01

Social cost-benefit analysis is a well-established method for guiding decisions about safety investments, particularly in situations in which it is possible to make accurate predictions of future performance. However, its direct applicability to situations involving large degrees of uncertainty is less obvious and this raises the question of the extent to which social cost-benefit analysis can provide a useful input to the decision framework that has been explicitly developed to deal with safety decisions in which uncertainty is a major factor, namely risk analysis. This is the main focus of the arguments developed in this paper. In particular, we provide new insights by examining the fundamentals of both approaches and our principal conclusion is that social cost-benefit analysis and risk analysis represent complementary input bases to the decision-making process, and even in the case of large uncertainties social cost-benefit analysis may provide very useful decision support. What is required is the establishment of a proper contextual framework which structures and gives adequate weight to the uncertainties. An application to the possibility of a robbery at a cash depot is examined as a practical example.

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

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

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.)

Putting Safety in the Frame

Directory of Open Access Journals (Sweden)

Valerie Jean O’Keeffe

2015-06-01

Full Text Available Current patient safety policy focuses nursing on patient care goals, often overriding nurses’ safety. Without understanding how nurses construct work health and safety (WHS, patient and nurse safety cannot be reconciled. Using ethnography, we examine social contexts of safety, studying 72 nurses across five Australian hospitals making decisions during patient encounters. In enacting safe practice, nurses used “frames” built from their contextual experiences to guide their behavior. Frames are produced by nurses, and they structure how nurses make sense of their work. Using thematic analysis, we identify four frames that inform nurses’ decisions about WHS: (a communicating builds knowledge, (b experiencing situations guides decisions, (c adapting procedures streamlines work, and (d team working promotes safe working. Nurses’ frames question current policy and practice by challenging how nurses’ safety is positioned relative to patient safety. Recognizing these frames can assist the design and implementation of effective WHS management.

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)

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

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

Root cause analysis of JCO accident based on decision-making model

International Nuclear Information System (INIS)

Kohda, Takehisa; Inoue, Koichi; Nojiri, Yoshihiko

2000-01-01

This paper discusses root causes of the JCO accident by considering the reasons why the workers made their decision to choose the illegal actions leading to a criticality accident. Analyzing their decision process compared with the normal decision process, the direct cause of their incorrect decision is estimated to be the lack of knowledge about the danger of nuclear materials and the criticality. Further, the lack of knowledge is considered to be due to organizational or environmental factors such as (a) the ignorance of safety by the overall JCO company which pursued low costs and high profit, (b) the JCO's custom and practice of modifying operational rules without permission, and (c) the JCO's inappropriate training or education where the criticality or its danger was not taught. All these background factors are related to the overconfidence of plant safety, a false trust that such a criticality accident will never occur at the plant. Since the recognition of the danger or risk of a system is considered to be the starting point for its safety management and operation, all information about the danger and safety should be correctly communicated to everyone related to the system. (author)

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.)

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

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．

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.

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

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

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...

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.

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.

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

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)

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.

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

Durable decision-making is central to the control of risks

International Nuclear Information System (INIS)

Zebroski, E.L.

1990-01-01

In several capacities the author has promoted the importance of risk analysis techniques as a rational path to the improved assurance of safety. The interest in decisions arose from the persistent observation of only moderate or minor impacts or benefits to practical operations from the availability of wall documented risk analysis studies for many nuclear units. The complexity and number of variables in decisions on matters of safety of large scale operations defies ordinary intuitive decision making. The structured decision process is not a panacea, but is often the practical tool of choise for managing complexity in an orderly way. Typical basic sources for decision techniques are listed in references. (orig.)

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

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

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

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

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.

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

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.

NKS/SOS-1 Seminar on Safety analysis. Report from a seminar held on 22-23 March 2000 Risø National Laboratory, Roskilde, DK

DEFF Research Database (Denmark)

The report describes presentations and discussions at a seminar held at Risø 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 multi-dimensional, which makes clarity...

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

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)

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…

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

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.)

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)

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

Safety case: An international perspective

International Nuclear Information System (INIS)

Pescatore, C.; Voinis, S.

2002-01-01

In recent years, it has become more and more evident that repository development will involve a number of stages punctuated by interdependent decisions on whether and how to move to the next stage. These decisions require a clear and traceable presentation of technical arguments that will help in giving confidence in the feasibility and safety of the proposed concept. The depth of understanding and technical information available to support decisions will vary from step to step. A safety case is a key item to support the decision to move to the next stage in repository development. Progress is noted, in the past decade, in the performance and safety assessment areas, particularly in the methodologies for repository system analysis. Progress is also observed regarding the understanding of the natural system and its characterisation, treatment of uncertainties, and modelling. Some areas are under active development, e.g. the area of scenario development and analysis. Finally, to increase confidence, rigorous quality assurance procedures need to be implemented, as well as the factoring of the contribution of R and D in underground research laboratories. The paper summarises the lessons learnt within relevant NEA initiatives as they evolved over the course of a decade and now allow a comprehensive view of what constitutes a safety case. (author)

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...

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

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.

Posiva's application for a decision in principle concerning a disposal facility for spent nuclear fuel. STUK's statement and preliminary safety appraisal

International Nuclear Information System (INIS)

Ruokola, E.

2000-03-01

In May 1999, Posiva Ltd submitted to the Government an application, pursuant to the Nuclear Energy Act, for a Decision in Principle on a disposal facility for spent nuclear fuel from the Finnish nuclear power plants. The Ministry of Trade and Industry requested the Radiation and Nuclear Safety Authority (STUK) to draw up a preliminary safety appraisal concerning the proposed disposal facility. In the beginning of this report, STUK's statement to the Ministry and Industry concerning the proposed disposal facility is given. In that statement, STUK concludes that the Decision in Principle is currently justified from the standpoint of safety. The statement is followed by a safety appraisal, where STUK deems, how the proposed disposal concept, site and facility comply with the safety requirements included in the Government's Decision (478/1999). STUK's preliminary safety appraisal was supported by contributions from a number of outside experts. A collective opinion by an international group of ten distinguished experts is appended to this report. (orig.)

Towards a Fuzzy Bayesian Network Based Approach for Safety Risk Analysis of Tunnel-Induced Pipeline Damage.

Science.gov (United States)

Zhang, Limao; Wu, Xianguo; Qin, Yawei; Skibniewski, Miroslaw J; Liu, Wenli

2016-02-01

Tunneling excavation is bound to produce significant disturbances to surrounding environments, and the tunnel-induced damage to adjacent underground buried pipelines is of considerable importance for geotechnical practice. A fuzzy Bayesian networks (FBNs) based approach for safety risk analysis is developed in this article with detailed step-by-step procedures, consisting of risk mechanism analysis, the FBN model establishment, fuzzification, FBN-based inference, defuzzification, and decision making. In accordance with the failure mechanism analysis, a tunnel-induced pipeline damage model is proposed to reveal the cause-effect relationships between the pipeline damage and its influential variables. In terms of the fuzzification process, an expert confidence indicator is proposed to reveal the reliability of the data when determining the fuzzy probability of occurrence of basic events, with both the judgment ability level and the subjectivity reliability level taken into account. By means of the fuzzy Bayesian inference, the approach proposed in this article is capable of calculating the probability distribution of potential safety risks and identifying the most likely potential causes of accidents under both prior knowledge and given evidence circumstances. A case concerning the safety analysis of underground buried pipelines adjacent to the construction of the Wuhan Yangtze River Tunnel is presented. The results demonstrate the feasibility of the proposed FBN approach and its application potential. The proposed approach can be used as a decision tool to provide support for safety assurance and management in tunnel construction, and thus increase the likelihood of a successful project in a complex project environment. © 2015 Society for Risk Analysis.

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

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)

Safety indicators: an efficient tool for a better safety

International Nuclear Information System (INIS)

Aufort, P.; Lars, R.

1993-01-01

Safety indicators based on the examination of the Operating Technical Specifications have been defined with the aim of following the in-operation safety level of French nuclear power plants. These safety indicators are operation feedback tools which permit the a posteriori justification and the adjustment of actual procedures. They would allow detection of an abnormal unavailability occurrence rate or a situation revealing a potential safety problem. So, data acquisition, processing, analysis and display software allowing trend analysis of these indicators has been developed so far as: a reflexion tool for the power plant operators about the safety instructions and the adjustment of preventive maintenance, and a help for decision making at a national level for the examination and the improvement of Operating Technical Specifications. This paper presents the objectives of these safety indicators, the processing tool associated, the preliminary results obtained and more elaborate processing of these indicators. These safety indicators may be very useful in framing probabilistic safety assessments. (author)

Applying the results of probablistic safety analysis of nuclear power plants: a survey of experience

International Nuclear Information System (INIS)

Andrews, W.B.; Herttrich, M.; Koeberlein, K.; Schwager.

1985-01-01

To date, discussions of the many different types of potential applications of PRA/PSA results and insights to safety-decision-making have been mainly theoretical. Various safety goals have been proposed as decision criteria. However, the discussion on the role of PRA/PSA and Safety Goals in safety-decision-making, especially in licensing, is controversial. A Working Group of the OECD Nuclear Energy Agency is completing a compilation and evaluation of real examples of past and present practical experience with the application of probabilistic methods in reactor safety decision-making, with the idea of developing a common understanding in this area. More than fifty different cases where PRA has influenced decision-making have been surveyed. These include, for example, regulatory changes, fixing of licensing requirements, plant specific modifications of design of operation, prioritization of safety issues and emergency planning. This feedback of experience - both positive and negative - with PRA/PSA applications is considered to provide guidance on how probabilistic approaches can be introduced into current safety practices, and on desirable future developments in probabilistic methods and specific PSA/PRA studies. Generic insights from the survey are given

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

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

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

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.

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

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....

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.

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

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.

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

Freight-train derailment rates for railroad safety and risk analysis.

Science.gov (United States)

Liu, Xiang; Rapik Saat, M; Barkan, Christopher P L

2017-01-01

Derailments are the most common type of train accident in the United States. They cause damage to infrastructure, rolling stock and lading, disrupt service, and have the potential to cause casualties, and harm the environment. Train safety and risk analysis relies on accurate assessment of derailment likelihood. Derailment rate - the number of derailments normalized by traffic exposure - is a useful statistic to estimate the likelihood of a derailment. Despite its importance, derailment rate analysis using multiple factors has not been previously developed. In this paper, we present an analysis of derailment rates on Class I railroad mainlines based on data from the U.S. Federal Railroad Administration and the major freight railroads. The point estimator and confidence interval of train and car derailment rates are developed by FRA track class, method of operation and annual traffic density. The analysis shows that signaled track with higher FRA track class and higher traffic density is associated with a lower derailment rate. The new accident rates have important implications for safety and risk management decisions, such as the routing of hazardous materials. Copyright © 2016 Elsevier Ltd. All rights reserved.

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)

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

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

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

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

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

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

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

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

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

Solid waste burial grounds interim safety analysis

International Nuclear Information System (INIS)

Saito, G.H.

1994-01-01

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

Solid waste burial grounds interim safety analysis

Energy Technology Data Exchange (ETDEWEB)

Saito, G.H.

1994-10-01

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

Safety 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

NASA System Safety Handbook. Volume 2: System Safety Concepts, Guidelines, and Implementation Examples

Science.gov (United States)

Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Feather, Martin; Rutledge, Peter; Sen, Dev; Youngblood, Robert

2015-01-01

This is the second of two volumes that collectively comprise the NASA System Safety Handbook. Volume 1 (NASASP-210-580) was prepared for the purpose of presenting the overall framework for System Safety and for providing the general concepts needed to implement the framework. Volume 2 provides guidance for implementing these concepts as an integral part of systems engineering and risk management. This guidance addresses the following functional areas: 1.The development of objectives that collectively define adequate safety for a system, and the safety requirements derived from these objectives that are levied on the system. 2.The conduct of system safety activities, performed to meet the safety requirements, with specific emphasis on the conduct of integrated safety analysis (ISA) as a fundamental means by which systems engineering and risk management decisions are risk-informed. 3.The development of a risk-informed safety case (RISC) at major milestone reviews to argue that the systems safety objectives are satisfied (and therefore that the system is adequately safe). 4.The evaluation of the RISC (including supporting evidence) using a defined set of evaluation criteria, to assess the veracity of the claims made therein in order to support risk acceptance decisions.

Analysis of safety impacts from external flooding using the risk-informed safety margin characterization (RISMC) Toolkit

International Nuclear Information System (INIS)

Smith, Curtis L.; Mandelli, Diego; Prescott, Steve

2015-01-01

The existing fleet of U.S. nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper demonstrates how Idaho National Laboratory (INL) researchers use the RISMC Toolkit to investigate complex nuclear plant phenomena using RAVEN and RELAP-7. The analysis focused on a highly relevant topic currently facing some nuclear power plants – specifically flooding issues. This research and development looked at challenges to a hypothetical pressurized water reactor, including: (1) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (2) earthquake induced station-blackout, and (3) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at INL. Using RAVEN, we were able to perform multiple RELAP-7 simulation runs by changing specific parts of the model in order to reflect specific aspects of different scenarios, including both the failure and recovery of critical components. The simulation employed traditional statistical tools (such as Monte-Carlo sampling) and more advanced machine-learning based algorithms to perform uncertainty quantification in order to understand changes in system performance and limitations as a consequence of power uprate. Qualitative and quantitative results obtained gave a detailed picture of the issues associated with potential accident scenarios. These types of

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

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

A comparison of workplace safety perceptions among financial decision-makers of medium- vs. large-size companies.

Science.gov (United States)

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

2011-01-01

This study, through a random national survey in the U.S., explored how corporate financial decision-makers perceive important workplace safety issues as a function of the size of the company for which they worked (medium- vs. large-size companies). Telephone surveys were conducted with 404 U.S. corporate financial decision-makers: 203 from medium-size companies and 201 from large companies. Results showed that the patterns of responding for participants from medium- and large-size companies were somewhat similar. The top-rated safety priorities in resource allocation reported by participants from both groups were overexertion, repetitive motion, and bodily reaction. They believed that there were direct and indirect costs associated with workplace injuries and for every dollar spent improving workplace safety, more than four dollars would be returned. They perceived the top benefits of an effective safety program to be predominately financial in nature - increased productivity and reduced costs - and the safety modification participants mentioned most often was to have more/better safety-focused training. However, more participants from large- than medium-size companies reported that "falling on the same level" was the major cause of workers' compensation loss, which is in line with industry loss data. Participants from large companies were more likely to see their safety programs as better than those of other companies in their industries, and those of medium-size companies were more likely to mention that there were no improvements needed for their companies. Copyright © 2009 Elsevier Ltd. All rights reserved.

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.

Putting Safety in the Frame: Nurses' Sensemaking at Work.

Science.gov (United States)

O'Keeffe, Valerie Jean; Thompson, Kirrilly Rebecca; Tuckey, Michelle Rae; Blewett, Verna Lesley

2015-01-01

Current patient safety policy focuses nursing on patient care goals, often overriding nurses' safety. Without understanding how nurses construct work health and safety (WHS), patient and nurse safety cannot be reconciled. Using ethnography, we examine social contexts of safety, studying 72 nurses across five Australian hospitals making decisions during patient encounters. In enacting safe practice, nurses used "frames" built from their contextual experiences to guide their behavior. Frames are produced by nurses, and they structure how nurses make sense of their work. Using thematic analysis, we identify four frames that inform nurses' decisions about WHS: (a) communicating builds knowledge, (b) experiencing situations guides decisions, (c) adapting procedures streamlines work, and (d) team working promotes safe working. Nurses' frames question current policy and practice by challenging how nurses' safety is positioned relative to patient safety. Recognizing these frames can assist the design and implementation of effective WHS management.

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....

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.

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

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

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.

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

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)

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.

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)

A life cycle analysis approach to D and D decision-making

International Nuclear Information System (INIS)

Yuracko, K.L.; Gresalfi, M.; Yerace, P.; Krstich, M.; Gerrick, D.

1998-05-01

This paper describes a life cycle analysis (LCA) approach that makes decontamination and decommissioning (D and D) of US Department of Energy facilities more efficient and more responsive to the concerns of the society. With the considerable complexity of D and D projects and their attendant environmental and health consequences, projects can no longer be designed based on engineering and economic criteria alone. Using the LCA D and D approach, the evaluation of material disposition alternatives explicitly includes environmental impacts, health and safety impacts, socioeconomic impacts, and stakeholder attitudes -- in addition to engineering and economic criteria. Multi-attribute decision analysis is used to take into consideration the uncertainties and value judgments that are an important part of all material disposition decisions. Use of the LCA D and D approach should lead to more appropriate selections of material disposition pathways and a decision-making process that is both understandable and defensible. The methodology and procedures of the LCA D and D approach are outlined and illustrated by an application of the approach at the Department of Energy's West Valley Demonstration Project. Specifically, LCA was used to aid decisions on disposition of soil and concrete from the Tank Pad D and D Project. A decision tree and the Pollution Prevention/Waste Minimization Users Guide for Environmental Restoration Projects were used to identify possible alternatives for disposition of the soil and concrete. Eight alternatives encompassing source reduction, segregation, treatment, and disposal were defined for disposition of the soil; two alternatives were identified for disposition of the concrete. Preliminary results suggest that segregation and treatment are advantageous in the disposition of both the soil and the concrete. This and other recent applications illustrate the strength and ease of application of the LCA D and D approach

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

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.

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

The Pashtun behavior economy an analysis of decision making in tribal society

OpenAIRE

Holton, Jeremy W.

2011-01-01

Approved for public release; distribution is unlimited. Little scholarship exists regarding the ways members of conflict societies think about the economic decisions they face, and what information they value as relevant to those decisions. The literature of the emerging field of behavior economics suggest that in uncertain environments, considerable weight may be given to identity and culture factors to make decisions that will affect personal safety, income prospects and self-fulfillment...

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.

Probabilistic safety assessment - regulatory perspective

International Nuclear Information System (INIS)

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

2002-01-01

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

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)

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)

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.

Risk-informed decision-making analysis for the electrical raceway fire barrier systems on a BWR-4 plant

International Nuclear Information System (INIS)

Wu, Ching-Hui; Lin, Tsu-Jen; Kao, Tsu-Mu; Chen, Chyn-Rong

2003-01-01

This paper describes a risk-informed decision-making approach used to resolve the fire barrier issue in a BWR-4 nuclear plant where Appendix R separation requirements cannot be met without installing additional fire protection features such as electrical raceway fire barrier system. The related risk measures in CDF (core damage frequency) and LERF (large early release frequency) of the fire barrier issue can be determined by calculating the difference in plant risks between various alternative cases and that met the requirement of the Appendix R. In some alternative cases, additional early-detection and fast-response fire suppression systems are suggested. In some other cases, cable re-routing of some improper layout of non-safety related cables are required. Sets of fire scenarios are re-evaluated more detailed by reviewing the cable damage impact for the BWR-4 plant. The fire hazard model, COMPBRM III-e, is used in this study and the dominant results in risk measures are benchmarked with the CFD code, FDS 2.0, to ensure that the risk impact of fire barrier is estimated accurately in the risk-informed decision making. The traditional deterministic qualitative methods, such as defense-in-depth, safety margin and post-fire safety shutdown capability are also proceeded. The value-impact analysis for proposed alternatives of fire wrapping required by Appendix R has been completed for technical basis of the exemption on Appendix R application. The outcome of the above analysis should be in compliance with the regulatory guidelines (RG) 1.174 and 1.189 for the applications in the risk-informed decision-making of the fire wrapping issues. (author)

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)

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...

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)

Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

International Nuclear Information System (INIS)

2016-01-01

This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication

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

International Nuclear Information System (INIS)

EL-Kafas, A.E.A.E.

1996-01-01

the purpose of the dissertation is to develop a 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 of 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 asses 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. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

Latent segmentation based count models: Analysis of bicycle safety in Montreal and Toronto.

Science.gov (United States)

Yasmin, Shamsunnahar; Eluru, Naveen

2016-10-01

The study contributes to literature on bicycle safety by building on the traditional count regression models to investigate factors affecting bicycle crashes at the Traffic Analysis Zone (TAZ) level. TAZ is a traffic related geographic entity which is most frequently used as spatial unit for macroscopic crash risk analysis. In conventional count models, the impact of exogenous factors is restricted to be the same across the entire region. However, it is possible that the influence of exogenous factors might vary across different TAZs. To accommodate for the potential variation in the impact of exogenous factors we formulate latent segmentation based count models. Specifically, we formulate and estimate latent segmentation based Poisson (LP) and latent segmentation based Negative Binomial (LNB) models to study bicycle crash counts. In our latent segmentation approach, we allow for more than two segments and also consider a large set of variables in segmentation and segment specific models. The formulated models are estimated using bicycle-motor vehicle crash data from the Island of Montreal and City of Toronto for the years 2006 through 2010. The TAZ level variables considered in our analysis include accessibility measures, exposure measures, sociodemographic characteristics, socioeconomic characteristics, road network characteristics and built environment. A policy analysis is also conducted to illustrate the applicability of the proposed model for planning purposes. This macro-level research would assist decision makers, transportation officials and community planners to make informed decisions to proactively improve bicycle safety - a prerequisite to promoting a culture of active transportation. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

Knowledge and perceived implementation of food safety risk analysis framework in Latin America and the Caribbean region.

Science.gov (United States)

Cherry, C; Mohr, A Hofelich; Lindsay, T; Diez-Gonzalez, F; Hueston, W; Sampedro, F

2014-12-01

Risk analysis is increasingly promoted as a tool to support science-based decisions regarding food safety. An online survey comprising 45 questions was used to gather information on the implementation of food safety risk analysis within the Latin American and Caribbean regions. Professionals working in food safety in academia, government, and private sectors in Latin American and Caribbean countries were contacted by email and surveyed to assess their individual knowledge of risk analysis and perceptions of its implementation in the region. From a total of 279 participants, 97% reported a familiarity with risk analysis concepts; however, fewer than 25% were able to correctly identify its key principles. The reported implementation of risk analysis among the different professional sectors was relatively low (46%). Participants from industries in countries with a long history of trade with the United States and the European Union, such as Mexico, Brazil, and Chile, reported perceptions of a higher degree of risk analysis implementation (56, 50, and 20%, respectively) than those from the rest of the countries, suggesting that commerce may be a driver for achieving higher food safety standards. Disagreement among respondents on the extent of the use of risk analysis in national food safety regulations was common, illustrating a systematic lack of understanding of the current regulatory status of the country. The results of this survey can be used to target further risk analysis training on selected sectors and countries.

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

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...

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...

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)

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

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

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

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)

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)

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

Methodology and development of instruments for the safety analysis of a nuclear reprocessing plant

International Nuclear Information System (INIS)

Markett, J.

1987-01-01

Characteristics and overlapping aspects in the elaboration of safety analyses for the nuclear and conventional units are presented. The current methods are presented and their limits of applicability characterized. The transferability of individual methods or their elements to the analysis of the reference plant of Wackersdorf is examined and the procedure for the systems analysis is determined. It is of great importance to prove that the essential kinds of incidents and possibilities of release with potential effects in the environment are completely identified. The incidents are divided into basic incidents, which are characterized by superior physical/chemical release mechanisms. An essential objective is to systematize the safety analysis and to summarize the presentation of results. Selection criteria are presented, which allow a limitation of the analysis to essential influencing parameters without removing aspects from the overall safety-relevant statement. Besides the selection criteria, instruments and mathematical models are explained with the help of which the representative and possible incidents covering all potential risks for all areas of the plant, systems and components can be selected. These design-basis accidents (criticality, self-heating, fire, explosion, leakages, earth quakes) are decisive for the determination of potential damaging effects in the environment and thus for the overall statement on the licensability. (orig./HP) [de

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

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.

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

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

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. ...

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.

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.

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

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

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

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

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.

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.

Probabilistic Safety Analysis of High Speed and Conventional Lines Using Bayesian Networks

Energy Technology Data Exchange (ETDEWEB)

Grande Andrade, Z.; Castillo Ron, E.; O' Connor, A.; Nogal, M.

2016-07-01

A Bayesian network approach is presented for probabilistic safety analysis (PSA) of railway lines. The idea consists of identifying and reproducing all the elements that the train encounters when circulating along a railway line, such as light and speed limit signals, tunnel or viaduct entries or exits, cuttings and embankments, acoustic sounds received in the cabin, curves, switches, etc. In addition, since the human error is very relevant for safety evaluation, the automatic train protection (ATP) systems and the driver behavior and its time evolution are modelled and taken into account to determine the probabilities of human errors. The nodes of the Bayesian network, their links and the associated probability tables are automatically constructed based on the line data that need to be carefully given. The conditional probability tables are reproduced by closed formulas, which facilitate the modelling and the sensitivity analysis. A sorted list of the most dangerous elements in the line is obtained, which permits making decisions about the line safety and programming maintenance operations in order to optimize them and reduce the maintenance costs substantially. The proposed methodology is illustrated by its application to several cases that include real lines such as the Palencia-Santander and the Dublin-Belfast lines. (Author)

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)

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

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.)

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.

Posiva's application for a decision in principle concerning a disposal facility for spent nuclear fuel. STUK's statement and preliminary safety appraisal

Energy Technology Data Exchange (ETDEWEB)

Ruokola, E. [ed.

2000-03-01

In May 1999, Posiva Ltd submitted to the Government an application, pursuant to the Nuclear Energy Act, for a Decision in Principle on a disposal facility for spent nuclear fuel from the Finnish nuclear power plants. The Ministry of Trade and Industry requested the Radiation and Nuclear Safety Authority (STUK) to draw up a preliminary safety appraisal concerning the proposed disposal facility. In the beginning of this report, STUK's statement to the Ministry and Industry concerning the proposed disposal facility is given. In that statement, STUK concludes that the Decision in Principle is currently justified from the standpoint of safety. The statement is followed by a safety appraisal, where STUK deems, how the proposed disposal concept, site and facility comply with the safety requirements included in the Government's Decision (478/1999). STUK's preliminary safety appraisal was supported by contributions from a number of outside experts. A collective opinion by an international group of ten distinguished experts is appended to this report. (orig.)

Nuclear power safety economics

International Nuclear Information System (INIS)

Legasov, V.A.; Demin, V.F.; Shevelev, Ya.V.

1984-01-01

The existing conceptual and methodical basis for the decision-making process insuring safety of the nuclear power and other (industrial and non-industrial) human activities is critically analyzed. Necessity of development a generalized economic safety analysis method (GESAM) is shown. Its purpose is justifying safety measures. Problems of GESAM development are considered including the problem of costing human risk. A number of suggestions on solving them are given. Using the discounting procedure in the assessment of risk or detriment caused by harmful impact on human health is substantiated. Examples of analyzing some safety systems in the nuclear power and other spheres of human activity are given

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

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

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.

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

Control Decisions for Flammable Gas Hazards in Waste Transfer Systems

International Nuclear Information System (INIS)

KRIPPS, L.J.

2000-01-01

This report describes the control decisions for flammable gas hazards in waste transfer systems (i.e., waste transfer piping and waste transfer-associated structures) made at control decision meetings on November 30, 1999a and April 19, 2000, and their basis. These control decisions, and the analyses that support them, will be documented in an amendment to the Final Safety Analysis Report (FSAR) (CHG 2000a) and Technical Safety Requirements (TSR) (CHG 2000b) to close the Flammable Gas Unreviewed Safety Question (USQ) (Bacon 1996 and Wagoner 1996). Following the Contractor Tier I review of the FSAR and TSR amendment, it will be submitted to the US. Department of Energy (DOE), Office of River Protection (ORP) for review and approval. The control decision meeting on November 30, 1999 to address flammable gas hazards in waste transfer systems followed the control decision process and the criteria for control decisions described in Section 3.3.1.5 of the FSAR. The control decision meeting agenda, attendance list, and introductory and background presentations are included in Attachments 1 through 4. The control decision discussions on existing and other possible controls for flammable gas hazards in waste transfer systems and the basis for selecting or not selecting specific controls are summarized in this report

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.

Road Safety Data, Collection, Transfer and Analysis DaCoTa. Deliverable 1.2: Road safety management investigation model and questionnaire.

NARCIS (Netherlands)

Dupont, H. Martensen, H. Papadimitriou, E. Yannis, G. Muhlrad, N. Jähi, H. Vallet, G. Giustiniani, G. Tripodi, A. Usami, D. Bax, C. Wijnen, W. Schöne, M.-L. Machata, K. Buttler, I. Zysinska, M. Talbot, R. Gitelman, V. & Hakkert, S. & Muhlrad, N. Gitelman, V. & Buttler, I. (Eds.)

2012-01-01

The aim of the DaCoTA Work Package 1 is to investigate road safety policy-making and management processes in Europe. In the Deliverables released previously, the Work Package 1 assessed the experts’ needs in terms of road safety knowledge, data and decision support tools (Deliverable 1.1/4.1), as

Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

Sensitivity analysis of model output - a step towards robust safety indicators?

International Nuclear Information System (INIS)

Broed, R.; Pereira, A.; Moberg, L.

2004-01-01

The protection of the environment from ionising radiation challenges the radioecological community with the issue of harmonising disparate safety indicators. These indicators should preferably cover the whole spectrum of model predictions on chemo-toxic and radiation impact of contaminants. In question is not only the protection of man and biota but also of abiotic systems. In many cases modelling will constitute the basis for an evaluation of potential impact. It is recognised that uncertainty and sensitivity analysis of model output will play an important role in the 'construction' of safety indicators that are robust, reliable and easy to explain to all groups of stakeholders including the general public. However, environmental models of transport of radionuclides have some extreme characteristics. They are, a) complex, b) non-linear, c) include a huge number of input parameters, d) input parameters are associated with large or very large uncertainties, e) parameters are often correlated to each other, f) uncertainties other than parameter-driven may be present in the modelling system, g) space variability and time-dependence of parameters are present, h) model predictions may cover geological time scales. Consequently, uncertainty and sensitivity analysis are non-trivial tasks, challenging the decision-maker when it comes to the interpretation of safety indicators or the application of regulatory criteria. In this work we use the IAEA model ISAM, to make a set of Monte Carlo calculations. The ISAM model includes several nuclides and decay chains, many compartments and variable parameters covering the range of nuclide migration pathways from the near field to the biosphere. The goal of our calculations is to make a global sensitivity analysis. After extracting the non-influential parameters, the M.C. calculations are repeated with those parameters frozen. Reducing the number of parameters to a few ones will simplify the interpretation of the results and the use

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

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

Active pedestrian safety by automatic braking and evasive steering

NARCIS (Netherlands)

Keller, C.; Dang, T.; Fritz, H.; Joos, A.; Rabe, C.; Gavrila, D.M.

2011-01-01

Active safety systems hold great potential for reducing accident frequency and severity by warning the driver and/or exerting automatic vehicle control ahead of crashes. This paper presents a novel active pedestrian safety system that combines sensing, situation analysis, decision making, and

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.

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)

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

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

Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities

International Nuclear Information System (INIS)

Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J.; Laub, T.W.

1992-06-01

This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10 -11 /yr to 10 -5 /yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10 -9 /yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution

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

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...

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.

RISMC advanced safety analysis project plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

International Nuclear Information System (INIS)

Szilard, Ronaldo H; Smith, Curtis L; Youngblood, Robert

2014-01-01

In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP 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. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (@@@why is this important?@@@) that will make the case for stakeholder's use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable @@use case@@@ demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

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.)

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)

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

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

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

European consumers and beef safety

DEFF Research Database (Denmark)

Van Wezemael, Lynn; Verbeke, Wim; Kügler, Jens Oliver

2010-01-01

European beef consumption has been gradually declining during the past decades, while consumers' concerns about beef safety have increased. This paper explores consumer perceptions of and interest in beef safety and beef safety information, and their role in beef safety assessment and the beef...... consumption decision making process. Eight focus group discussions were performed with a total of 65 beef consumers in four European countries. Content analysis revealed that European consumers experienced difficulties in the assessment of the safety of beef and beef products and adopted diverging uncertainty...... reduction strategies. These include the use of colour, labels, brands and indications of origin as cues signalling beef safety. In general, consumer trust in beef safety was relatively high, despite distrust in particular actors....

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

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.

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

NASA System Safety Handbook. Volume 1; System Safety Framework and Concepts for Implementation

Science.gov (United States)

Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Smith, Curtis; Stamatelatos, Michael; Youngblood, Robert

2011-01-01

basis but to consider measures of aggregate safety risk and to ensure wherever possible that there be quantitative measures for evaluating how effective the controls are in reducing these aggregate risks. The term aggregate risk, when used in this handbook, refers to the accumulation of risks from individual scenarios that lead to a shortfall in safety performance at a high level: e.g., an excessively high probability of loss of crew, loss of mission, planetary contamination, etc. Without aggregated quantitative measures such as these, it is not reasonable to expect that safety has been optimized with respect to other technical and programmatic objectives. At the same time, it is fully recognized that not all sources of risk are amenable to precise quantitative analysis and that the use of qualitative approaches and bounding estimates may be appropriate for those risk sources. Second, the handbook stresses the necessity of developing confidence that the controls derived for the purpose of achieving system safety not only handle risks that have been identified and properly characterized but also provide a general, more holistic means for protecting against unidentified or uncharacterized risks. For example, while it is not possible to be assured that all credible causes of risk have been identified, there are defenses that can provide protection against broad categories of risks and thereby increase the chances that individual causes are contained. Third, the handbook strives at all times to treat uncertainties as an integral aspect of risk and as a part of making decisions. The term "uncertainty" here does not refer to an actuarial type of data analysis, but rather to a characterization of our state of knowledge regarding results from logical and physical models that approximate reality. Uncertainty analysis finds how the output parameters of the models are related to plausible variations in the input parameters and in the modeling assumptions. The evaluation of

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.

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

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)

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)

The Best Path Analysis in Military Highway Transport Based on DEA and Multiobjective Fuzzy Decision-Making

Directory of Open Access Journals (Sweden)

Wu Juan

2014-01-01

Full Text Available Military transport path selection directly affects the transport speed, efficiency, and safety. To a certain degree, the results of the path selection determine success or failure of the war situation. The purpose of this paper is to propose a model based on DEA (data envelopment analysis and multiobjective fuzzy decision-making for path selection. The path decision set is established according to a search algorithm based on overlapping section punishment. Considering the influence of various fuzzy factors, the model of optimal path is constructed based on DEA and multitarget fuzzy decision-making theory, where travel time, transport risk, quick response capability, and transport cost constitute the evaluation target set. A reasonable path set can be calculated and sorted according to the comprehensive scores of the paths. The numerical results show that the model and the related algorithms are effective for path selection of military transport.

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

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

Control Decisions for Flammable Gas Hazards in Double Contained Receiver Tanks (DCRTs)

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2000-06-28

This report describes the control decisions for flammable gas hazards in double-contained receiver tanks (DCRTs) made at control decision meetings on November 16, 17, and 18, 1999, on April 19,2000, and on May 10,2000, and their basis. These control decisions, and the analyses that support them, will be documented in an amendment to the Final Safety Analysis Report (FSAR) (CHG 2000a) and Technical Safety Requirements (TSR) (CHG 2000b) to close the Flammable Gas Unreviewed Safety Question (USQ) (Bacon 1996 and Wagoner 1996) for DCRTs. Following the contractor Tier I review of the FSAR and TSR amendment, it will be submitted to the U.S. Department of Energy (DOE), Office of River Protection (ORP) for review and approval.

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.

Risk-Informed Safety Margin Characterization (RISMC): Integrated Treatment of Aleatory and Epistemic Uncertainty in Safety Analysis

International Nuclear Information System (INIS)

Youngblood, R.W.

2010-01-01

The concept of 'margin' has a long history in nuclear licensing and in the codification of good engineering practices. However, some traditional applications of 'margin' have been carried out for surrogate scenarios (such as design basis scenarios), without regard to the actual frequencies of those scenarios, and have been carried out with in a systematically conservative fashion. This means that the effectiveness of the application of the margin concept is determined in part by the original choice of surrogates, and is limited in any case by the degree of conservatism imposed on the evaluation. In the RISMC project, which is part of the Department of Energy's 'Light Water Reactor Sustainability Program' (LWRSP), we are developing a risk-informed characterization of safety margin. Beginning with the traditional discussion of 'margin' in terms of a 'load' (a physical challenge to system or component function) and a 'capacity' (the capability of that system or component to accommodate the challenge), we are developing the capability to characterize probabilistic load and capacity spectra, reflecting both aleatory and epistemic uncertainty in system response. For example, the probabilistic load spectrum will reflect the frequency of challenges of a particular severity. Such a characterization is required if decision-making is to be informed optimally. However, in order to enable the quantification of probabilistic load spectra, existing analysis capability needs to be extended. Accordingly, the INL is working on a next-generation safety analysis capability whose design will allow for much more efficient parameter uncertainty analysis, and will enable a much better integration of reliability-related and phenomenology-related aspects of margin.

A graded approach to safety documentation at processing facilities

International Nuclear Information System (INIS)

Cowen, M.L.

1992-01-01

Westinghouse Savannah River Company (WSRC) has over 40 major Safety Analysis Reports (SARs) in preparation for non-reactor facilities. These facilities include nuclear material production facilities, waste management facilities, support laboratories and environmental remediation facilities. The SARs for these various projects encompass hazard levels from High to Low, and mission times from startup, through operation, to shutdown. All of these efforts are competing for scarce resources, and therefore some mechanism is required for balancing the documentation requirements. Three of the key variables useful for the decision making process are Depth of Safety Analysis, Urgency of Safety Analysis, and Resource Availability. This report discusses safety documentation at processing facilities

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

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

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

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)

A formal safety analysis for PLC software-based safety critical system using Z

International Nuclear Information System (INIS)

Koh, Jung Soo

1997-02-01

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 formal 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. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

SLUDGE TREATMENT PROJECT KOP CONCEPTUAL DESIGN CONTROL DECISION REPORT

International Nuclear Information System (INIS)

Carro, C.A.

2010-01-01

This control decision addresses the Knock-Out Pot (KOP) Disposition KOP Processing System (KPS) conceptual design. The KPS functions to (1) retrieve KOP material from canisters, (2) remove particles less than 600 (micro)m in size and low density materials from the KOP material, (3) load the KOP material into Multi-Canister Overpack (MCO) baskets, and (4) stage the MCO baskets for subsequent loading into MCOs. Hazard and accident analyses of the KPS conceptual design have been performed to incorporate safety into the design process. The hazard analysis is documented in PRC-STP-00098, Knock-Out Pot Disposition Project Conceptual Design Hazard Analysis. The accident analysis is documented in PRC-STP-CN-N-00167, Knock-Out Pot Disposition Sub-Project Canister Over Lift Accident Analysis. Based on the results of these analyses, and analyses performed in support of MCO transportation and MCO processing and storage activities at the Cold Vacuum Drying Facility (CVDF) and Canister Storage Building (CSB), control decision meetings were held to determine the controls required to protect onsite and offsite receptors and facility workers. At the conceptual design stage, these controls are primarily defined by their safety functions. Safety significant structures, systems, and components (SSCs) that could provide the identified safety functions have been selected for the conceptual design. It is anticipated that some safety SSCs identified herein will be reclassified based on hazard and accident analyses performed in support of preliminary and detailed design.

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.

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

An overview-probabilistic safety analysis for research reactors

International Nuclear Information System (INIS)

Liu Jinlin; Peng Changhong

2015-01-01

For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

Computerized operator decision aids

International Nuclear Information System (INIS)

Long, A.B.

1984-01-01

This article explores the potential benefits associated with the use of computers in nuclear plants by the operating crew as an aid in making decisions. Pertinent findings are presented from recently completed projects to establish the context in which operating decisions have to be made. Key factors influencing the decision-making process itself are also identified. Safety parameter display systems, which are being implemented in various forms by the nuclear industry, are described within the context of decision making. In addition, relevant worldwide research and development activities are examined as potential enhancements to computerized operator decision aids to further improve plant safety and availability

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

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

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

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.)

Probabilistic safety analysis procedures guide, Sections 8-12. Volume 2, Rev. 1

International Nuclear Information System (INIS)

McCann, M.; Reed, J.; Ruger, C.; Shiu, K.; Teichmann, T.; Unione, A.; Youngblood, R.

1985-08-01

A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. 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 issues affecting reactor safety. The first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. This second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, 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. For external events, more explicit guidance is given

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

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

The affect heuristic in occupational safety.

Science.gov (United States)

Savadori, Lucia; Caovilla, Jessica; Zaniboni, Sara; Fraccaroli, Franco

2015-07-08

The affect heuristic is a rule of thumb according to which, in the process of making a judgment or decision, people use affect as a cue. If a stimulus elicits positive affect then risks associated to that stimulus are viewed as low and benefits as high; conversely, if the stimulus elicits negative affect, then risks are perceived as high and benefits as low. The basic tenet of this study is that affect heuristic guides worker's judgment and decision making in a risk situation. The more the worker likes her/his organization the less she/he will perceive the risks as high. A sample of 115 employers and 65 employees working in small family agricultural businesses completed a questionnaire measuring perceived safety costs, psychological safety climate, affective commitment and safety compliance. A multi-sample structural analysis supported the thesis that safety compliance can be explained through an affect-based heuristic reasoning, but only for employers. Positive affective commitment towards their family business reduced employers' compliance with safety procedures by increasing the perceived cost of implementing them.

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)

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

Operational readiness decisions at nuclear power plants. Which factors influence the decisions?

International Nuclear Information System (INIS)

Kecklund, Lena; Petterson, Sara

2007-11-01

The purpose of this project has been to propose a model for how operational readiness decisions are made and to identify important factors influencing these decisions. The project has also studied the support from the management system for decision making, and made a comparison to how decisions are made in practice. This is mainly an explorative study, but it also deals with relevant research and theories about decision making. The project consists of several parts. The first part is composed of descriptions of important notations and terms, and a summary of relevant research about decision making and its relation to the management system. The project proposes a model for the decision making process. The second part consists of analyses of reports from SKI about operational readiness decisions. The last part is a case study at a nuclear power plant. The case study describes the support from work method theories at the nuclear power plant to the decision maker. Decision makers with different roles in the safety management system were interviewed to give a description of the decision making process and of factors influencing the decisions made in practice. The case study also consists of an analysis of decisions in some real events at the nuclear power plant, as well as of making interviews in connection with these. To sum up, this report presents a model for the decision process and describes the work method theories that support the different parts in the process, how the different parts are applied in practice and circumstances that influence the decision process. The results of the project give an understanding for decision making in operational readiness decisions and the factors that influence the decision. The results are meant to be used as a basis for further studies in other nuclear power plants. The results indicate that the decision process is facilitated if there are clear criteria and work methods, if the work methods are well established and if the

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)

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

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

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...

29 CFR 1905.41 - Summary decision.

Science.gov (United States)

2010-07-01

... OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 Summary Decisions § 1905.41 Summary decision. (a) No genuine issue... 29 Labor 5 2010-07-01 2010-07-01 false Summary decision. 1905.41 Section 1905.41 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RULES OF...

Patient-Centered Robot-Aided Passive Neurorehabilitation Exercise Based on Safety-Motion Decision-Making Mechanism

Directory of Open Access Journals (Sweden)

Lizheng Pan

2017-01-01

Full Text Available Safety is one of the crucial issues for robot-aided neurorehabilitation exercise. When it comes to the passive rehabilitation training for stroke patients, the existing control strategies are usually just based on position control to carry out the training, and the patient is out of the controller. However, to some extent, the patient should be taken as a “cooperator” of the training activity, and the movement speed and range of the training movement should be dynamically regulated according to the internal or external state of the subject, just as what the therapist does in clinical therapy. This research presents a novel motion control strategy for patient-centered robot-aided passive neurorehabilitation exercise from the point of the safety. The safety-motion decision-making mechanism is developed to online observe and assess the physical state of training impaired-limb and motion performances and regulate the training parameters (motion speed and training rage, ensuring the safety of the supplied rehabilitation exercise. Meanwhile, position-based impedance control is employed to realize the trajectory tracking motion with interactive compliance. Functional experiments and clinical experiments are investigated with a healthy adult and four recruited stroke patients, respectively. The two types of experimental results demonstrate that the suggested control strategy not only serves with safety-motion training but also presents rehabilitation efficacy.

A Framework for Assessment of Aviation Safety Technology Portfolios

Science.gov (United States)

Jones, Sharon M.; Reveley, Mary S.

2014-01-01

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

PROFIT SENSITIVITY IN THE DECISION - MAKING PROCESS

Directory of Open Access Journals (Sweden)

Dimi Ofilean

2014-09-01

Full Text Available Projections on the profitability of an entity is a prerequisite impact assessment of implementing various management strategies. The literature did not include a model sensitivity analysis in terms of profit margin of safety modification and safety coefficient. This article aims to explicit solutions for identifying the factors that influence the sensitivity of profit, the proposed analytical models to change the margin of safety (physical and value and coefficient of safety. The model allows the determination of limits that can increase or decrease sales costs so that the company remains profitable, ie to be able to maintain an adequate level of profit. This analysis allows knowing the influence of each factor in the evolution of the profitability of the entity, allowing managers to adopt the right decisions based on the importance of the influence of the analysis results of the entity. To facilitate understanding of the proposed analytical model is presented a case study.

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...

Study Of Safety Management By Using Gis In Coimbatore

Directory of Open Access Journals (Sweden)

S. Kanchana

2015-08-01

Full Text Available The safety management is very important in the process of construction .The traditional methods of construction safety control cannot meet the construction of big project. To ensure the safety of construction and reduce accidents in the process of construction the current situation and problems we face in construction safety management should be studied first. And then the project risk warning mechanism based on the GIS is constructed according to the problems we faced to achieve visual monitoring and warning of construction safety risk management and to provide decision support for construction. This project aims to develop a web-based spatial decision support system model for proactive health and safety management in linear construction projects. 5 Currently health and safety management is usually performed reactively instead of proactive management since hazard identification and risk assessment is mostly performed on paper based documents that are not effectively used at site. An information system relates to a chain of operations lead to planning the observation and collection of data to storage and analysis of data to the use of derived information in decision-making processes. To create a web-based free and open sourced GIS that can work with different data formats by exchanging and presenting data as a real-time map on web.

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.)

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.

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.

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.

Optimization method concerning target conflicts between safety aspects and occupational safety aspects in nuclear power plant operations

International Nuclear Information System (INIS)

Mueller, W.

1991-01-01

The simplified cost-benefit analysis has not been considered for applications in nuclear engineering with complex decisions between safety aspects and occupational safety aspects. The extended cost-benefit analysis encounters problems with non-monetary criteria. Solutions are in sight, however with a subjective element. A major problem in implementing the method is the psychological barrier as against an evaluation of human life. The multi-attribute utility analysis overcomes the difficulties of the extended cost-benefit analysis, however, it also creates new problems on account of the complicated construction of the utility functions. The problems are solved most elegantly with the multi-criteria outranking analysis, the only disadvantage possibly being less transparency at first sight. (orig./HP) [de

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

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

An LWR design decision Methodology

International Nuclear Information System (INIS)

Leahy, T.J.; Rees, D.C.; Young, J.

1982-01-01

While all parties involved in nuclear plant regulation endeavor to make decisions which optimize the considerations of plant safety and financial impacts, these decisions are generally made without the benefit of a systematic and rigorous approach to the questions confronting the decision makers. A Design Decision Methodology has been developed which provides such a systematic approach. By employing this methodology, which makes use of currently accepted probabilistic risk assessment techniques and cost estimation, informed decisions may be made against a background of comparisons between the relative levels of safety and costs associated with various design alternatives

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)

Operational readiness decisions at nuclear power plants - part 2. Which factors influence the decisions?

International Nuclear Information System (INIS)

Kecklund, Lena; Petterson, Sara

2008-04-01

The first report contained a summary of relevant research of decision making, a case study at Ringhals power plant and an analysis of some real cases of operational readiness decisions. In this report two case studies in the Swedish power plants, OKG and Forsmark are presented. The case study description consists of three parts; a description of the support from the management system for the decision making process, interviews with decision makers and an analysis of real cases of operational readiness decisions. The purpose of the project has been to increase the understanding of the decision process in operational readiness decisions as well as the support given from the management system and what factors influence the decisions. From a general point of view the circumstances where the decision must be taken varies, but situations and events that lead to questioning of the operational readiness are often easy to identify. There are often support documents such as procedures, rules and technical documents which specify operational limitations which give explicit decision criteria. These decisions are easy. When needed colleagues can be consulted for support. In unclear situations and/or when the technical criteria is not clear, e.g. when the rules and regulations are vague or even in conflict or when it is not evident that you need to question the operational readiness, the decision is more difficult to make. The results from the study shows that such decisions in general are not made by the shift crew manager but handed over to the next management level. The decision making process differs between the power plants. At one of the power plants the decision process is organised in specific meetings where decision made are reviewed by the next higher management level. At another plant the decisions are often made in groups or in consultation with colleagues. The management system makes a distinction between decisions made in consultation and when decisions already

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

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)

Causal Relationship Analysis of the Patient Safety Culture Based on Safety Attitudes Questionnaire in Taiwan

Science.gov (United States)

Zeng, Pei-Shan; Huang, Chih-Hsuan

2018-01-01

This study uses the decision-making trial and evaluation laboratory method to identify critical dimensions of the safety attitudes questionnaire in Taiwan in order to improve the patient safety culture from experts' viewpoints. Teamwork climate, stress recognition, and perceptions of management are three causal dimensions, while safety climate, job satisfaction, and working conditions are receiving dimensions. In practice, improvements on effect-based dimensions might receive little effects when a great amount of efforts have been invested. In contrast, improving a causal dimension not only improves itself but also results in better performance of other dimension(s) directly affected by this particular dimension. Teamwork climate and perceptions of management are found to be the most critical dimensions because they are both causal dimensions and have significant influences on four dimensions apiece. It is worth to note that job satisfaction is the only dimension affected by the other dimensions. In order to effectively enhance the patient safety culture for healthcare organizations, teamwork climate, and perceptions of management should be closely monitored. PMID:29686825

Causal Relationship Analysis of the Patient Safety Culture Based on Safety Attitudes Questionnaire in Taiwan

Directory of Open Access Journals (Sweden)

Yii-Ching Lee

2018-01-01

Full Text Available This study uses the decision-making trial and evaluation laboratory method to identify critical dimensions of the safety attitudes questionnaire in Taiwan in order to improve the patient safety culture from experts’ viewpoints. Teamwork climate, stress recognition, and perceptions of management are three causal dimensions, while safety climate, job satisfaction, and working conditions are receiving dimensions. In practice, improvements on effect-based dimensions might receive little effects when a great amount of efforts have been invested. In contrast, improving a causal dimension not only improves itself but also results in better performance of other dimension(s directly affected by this particular dimension. Teamwork climate and perceptions of management are found to be the most critical dimensions because they are both causal dimensions and have significant influences on four dimensions apiece. It is worth to note that job satisfaction is the only dimension affected by the other dimensions. In order to effectively enhance the patient safety culture for healthcare organizations, teamwork climate, and perceptions of management should be closely monitored.

Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

2001-02-01

This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

International Nuclear Information System (INIS)

Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

2001-02-01

This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

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

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.

Analysis of international approaches which are used at development of theoperational safety performance indicators

International Nuclear Information System (INIS)

Lyigots'kij, O.Yi.; Nosovs'kij, A.V.; Chemeris, Yi.O.

2009-01-01

Description of international approaches and experience of the use of theoperational safety performance indicators system is provided for estimationof current status and making a decision on corrections in the operationpractice. The state of development of the operational safety performanceindicators system by the operating organization is overviewed. Thepossibility of application of international approaches during development ofthe integral safety performance indicators system is analyzed. Aims and tasksof future researches are formulated in relation to development of theintegral safety performance indicators system.

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...

Status of generic actions items and safety analysis system of PHWR

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Min, Byung Joo

2001-05-01

This report described the review results of a GAIs(Generic Action Item) currently issued on safety analysis of PHWR(Pressurized Heavy Water Reactor) and the research activities and positions to solve the GAIs in each country which possess PHWRs. eviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc.. were described. From the present review report, it is intended to establish the CANDU safety analysis system by providing the better understandings and development plans for the safety analysis of PHWR. esults.

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.

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.

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

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.)

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)

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

PWR core safety analysis with 3-dimensional methods

International Nuclear Information System (INIS)

Gensler, A.; Kühnel, K.; Kuch, S.

2015-01-01

Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

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

Mathematical Safety Assessment Approaches for Thermal Power Plants

Directory of Open Access Journals (Sweden)

Zong-Xiao Yang

2014-01-01

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

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.

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.

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

Shutdown Safety in NEK

International Nuclear Information System (INIS)

Gluhak, Mario; Senegovic, Marko

2014-01-01

Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity control, continuity of core cooling, and integrity of fission product barriers are valued as essential, distinguishing attributes of nuclear station work environment'. NEK has recognized the importance of 'defence in depth'Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity

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.

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)

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

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

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

Waste Transfer Leaks Control Decision Record

International Nuclear Information System (INIS)

RYAN, G.W.

2000-01-01

Control decision meetings for Waste Transfer Leaks were held on April 24,25,26, and 27, 2000. The agenda for the control decision meetings is included in Appendix A, and attendee lists are included in Appendix B. The purpose of the control decision meetings was to review and revise previously selected controls for the prevention or mitigation of waste transfer leak accidents. Re-evaluation of the controls is warranted due to revisions in the hazard and accident analysis for these Tank Farm events. In particular, calculated radiological consequences are significantly reduced from those currently reported in the Final Safety Analysis Report (FSAR). Revised hazard and accident analysis and a revised control recommendation will be reflected in an Authorization Basis Amendment to be submitted at the Department of Energy, Office of River Protection's (ORP's) request by June 30, 2000 to satisfy ORP Performance Incentive (PI) 2.1.1, Revision 1, ''Authorization Basis Management Process Efficiency Improvement''. The scope of the control decision meetings was to address all waste transfer leak-related hazardous conditions identified in the Tank Farm hazard analysis database, excluding those associated with the use of the Replacement Cross-Site Transfer System (RCSTS) slurry line and sluicing of Tank 241-C-106, which is addressed in FSAR Addendum 1. The scope of this control decision process does include future waste feed delivery waste transfer operations

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.

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.

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

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

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.

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.)

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 ...

Issues affecting advanced passive light-water reactor safety analysis