Potential effects of organizational uncertainty on safety

International Nuclear Information System (INIS)

Durbin, N.E.; Lekberg, A.; Melber, B.D.

2001-12-01

When organizations face significant change - reorganization, mergers, acquisitions, down sizing, plant closures or decommissioning - both the organizations and the workers in those organizations experience significant uncertainty about the future. This uncertainty affects the organization and the people working in the organization - adversely affecting morale, reducing concentration on safe operations, and resulting in the loss of key staff. Hence, organizations, particularly those using high risk technologies, which are facing significant change need to consider and plan for the effects of organizational uncertainty on safety - as well as planning for other consequences of change - technical, economic, emotional, and productivity related. This paper reviews some of what is known about the effects of uncertainty on organizations and individuals, discusses the potential consequences of uncertainty on organizational and individual behavior, and presents some of the implications for safety professionals

Dealing with uncertainty arising out of probabilistic risk assessment

International Nuclear Information System (INIS)

Solomon, K.A.; Kastenberg, W.E.; Nelson, P.F.

1984-03-01

In addressing the area of safety goal implementation, the question of uncertainty arises. This report suggests that the Nuclear Regulatory Commission (NRC) should examine how other regulatory organizations have addressed the issue. Several examples are given from the chemical industry, and comparisons are made to nuclear power risks. Recommendations are made as to various considerations that the NRC should require in probabilistic risk assessments in order to properly treat uncertainties in the implementation of the safety goal policy. 40 references, 7 figures, 5 tables

Safety assessment as basis for the decision making process

International Nuclear Information System (INIS)

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

2005-01-01

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

Robust Adaptation? Assessing the sensitivity of safety margins in flood defences to uncertainty in future simulations - a case study from Ireland.

Science.gov (United States)

Murphy, Conor; Bastola, Satish; Sweeney, John

2013-04-01

Climate change impact and adaptation assessments have traditionally adopted a 'top-down' scenario based approach, where information from different Global Climate Models (GCMs) and emission scenarios are employed to develop impacts led adaptation strategies. Due to the tradeoffs in the computational cost and need to include a wide range of GCMs for fuller characterization of uncertainties, scenarios are better used for sensitivity testing and adaptation options appraisal. One common approach to adaptation that has been defined as robust is the use of safety margins. In this work the sensitivity of safety margins that have been adopted by the agency responsible for flood risk management in Ireland, to the uncertainty in future projections are examined. The sensitivity of fluvial flood risk to climate change is assessed for four Irish catchments using a large number of GCMs (17) forced with three emissions scenarios (SRES A1B, A2, B1) as input to four hydrological models. Both uncertainty within and between hydrological models is assessed using the GLUE framework. Regionalisation is achieved using a change factor method to infer changes in the parameters of a weather generator using monthly output from the GCMs, while flood frequency analysis is conducted using the method of probability weighted moments to fit the Generalised Extreme Value distribution to ~20,000 annual maxima series. The sensitivity of design margins to the uncertainty space considered is visualised using risk response surfaces. The hydrological sensitivity is measured as the percentage change in flood peak for specified recurrence intervals. Results indicate that there is a considerable residual risk associated with allowances of +20% when uncertainties are accounted for and that the risk of exceedence of design allowances is greatest for more extreme, low frequency events with considerable implication for critical infrastructure, e.g., culverts, bridges, flood defences whose designs are normally

Quantification of Safety-Critical Software Test Uncertainty

International Nuclear Information System (INIS)

Khalaquzzaman, M.; Cho, Jaehyun; Lee, Seung Jun; Jung, Wondea

2015-01-01

The method, conservatively assumes that the failure probability of a software for the untested inputs is 1, and the failure probability turns in 0 for successful testing of all test cases. However, in reality the chance of failure exists due to the test uncertainty. Some studies have been carried out to identify the test attributes that affect the test quality. Cao discussed the testing effort, testing coverage, and testing environment. Management of the test uncertainties was discussed in. In this study, the test uncertainty has been considered to estimate the software failure probability because the software testing process is considered to be inherently uncertain. A reliability estimation of software is very important for a probabilistic safety analysis of a digital safety critical system of NPPs. This study focused on the estimation of the probability of a software failure that considers the uncertainty in software testing. In our study, BBN has been employed as an example model for software test uncertainty quantification. Although it can be argued that the direct expert elicitation of test uncertainty is much simpler than BBN estimation, however the BBN approach provides more insights and a basis for uncertainty estimation

Generally Recognized as Safe: Uncertainty Surrounding E-Cigarette Flavoring Safety

Directory of Open Access Journals (Sweden)

Clara G. Sears

2017-10-01

Full Text Available Despite scientific uncertainty regarding the relative safety of inhaling e-cigarette aerosol and flavorings, some consumers regard the U.S. Food and Drug Administration’s “generally recognized as safe” (GRAS designation as evidence of flavoring safety. In this study, we assessed how college students’ perceptions of e-cigarette flavoring safety are related to understanding of the GRAS designation. During spring 2017, an online questionnaire was administered to college students. Chi-square p-values and multivariable logistic regression were employed to compare perceptions among participants considering e-cigarette flavorings as safe and those considering e-cigarette flavorings to be unsafe. The total sample size was 567 participants. Only 22% knew that GRAS designation meant that a product is safe to ingest, not inhale, inject, or use topically. Of participants who considered flavorings to be GRAS, the majority recognized that the designation meant a product is safe to ingest but also considered it safe to inhale. Although scientific uncertainty on the overall safety of flavorings in e-cigarettes remains, health messaging can educate the public about the GRAS designation and its irrelevance to e-cigarette safety.

Uncertainties and reliability theories for reactor safety

International Nuclear Information System (INIS)

Veneziano, D.

1975-01-01

What makes the safety problem of nuclear reactors particularly challenging is the demand for high levels of reliability and the limitation of statistical information. The latter is an unfortunate circumstance, which forces deductive theories of reliability to use models and parameter values with weak factual support. The uncertainty about probabilistic models and parameters which are inferred from limited statistical evidence can be quantified and incorporated rationally into inductive theories of reliability. In such theories, the starting point is the information actually available, as opposed to an estimated probabilistic model. But, while the necessity of introducing inductive uncertainty into reliability theories has been recognized by many authors, no satisfactory inductive theory is presently available. The paper presents: a classification of uncertainties and of reliability models for reactor safety; a general methodology to include these uncertainties into reliability analysis; a discussion about the relative advantages and the limitations of various reliability theories (specifically, of inductive and deductive, parametric and nonparametric, second-moment and full-distribution theories). For example, it is shown that second-moment theories, which were originally suggested to cope with the scarcity of data, and which have been proposed recently for the safety analysis of secondary containment vessels, are the least capable of incorporating statistical uncertainty. The focus is on reliability models for external threats (seismic accelerations and tornadoes). As an application example, the effect of statistical uncertainty on seismic risk is studied using parametric full-distribution models

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)

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.)

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

International Nuclear Information System (INIS)

Ebashi, Takeshi; Kaku, Kenichi; Ishiguro, Katsuhiko

2012-01-01

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

Safety Goal, Multi-unit Risk and PSA Uncertainty

Energy Technology Data Exchange (ETDEWEB)

Yang, Joon-Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The safety goal is an answer of each country to the question 'How safe is safe enough?'. Table 1 shows some examples of the safety goal. However, many countries including Korea do not have the official safety goal for NPPs up to now since the establishment of safety goal is not just a technical issue but a very complex socio-technical issue. In establishing the safety goal for nuclear facilities, we have to consider various factors including not only technical aspects but also social, cultural ones. Recently, Korea is trying to establish the official safety goal. In this paper, we will review the relationship between the safety goal and Probabilistic Safety Assessment (PSA). We will also address some important technical issues to be considered in establishing the safety goal for NPPs from PSA point of view, i.e. a multi-unit risk issue and the uncertainty of PSA. In this paper, we reviewed some issues related to the safety goal and PSA. We believe that the safety goal is to be established in Korea considering the multi-unit risk. In addition, the relationship between the safety goal and PSA should be also defined clearly since PSA is the only way to answer to the question 'How safe is safe enough?'.

Epistemic and aleatory uncertainties in integrated deterministic and probabilistic safety assessment: Tradeoff between accuracy and accident simulations

International Nuclear Information System (INIS)

Karanki, D.R.; Rahman, S.; Dang, V.N.; Zerkak, O.

2017-01-01

The coupling of plant simulation models and stochastic models representing failure events in Dynamic Event Trees (DET) is a framework used to model the dynamic interactions among physical processes, equipment failures, and operator responses. The integration of physical and stochastic models may additionally enhance the treatment of uncertainties. Probabilistic Safety Assessments as currently implemented propagate the (epistemic) uncertainties in failure probabilities, rates, and frequencies; while the uncertainties in the physical model (parameters) are not propagated. The coupling of deterministic (physical) and probabilistic models in integrated simulations such as DET allows both types of uncertainties to be considered. However, integrated accident simulations with epistemic uncertainties will challenge even today's high performance computing infrastructure, especially for simulations of inherently complex nuclear or chemical plants. Conversely, intentionally limiting computations for practical reasons would compromise accuracy of results. This work investigates how to tradeoff accuracy and computations to quantify risk in light of both uncertainties and accident dynamics. A simple depleting tank problem that can be solved analytically is considered to examine the adequacy of a discrete DET approach. The results show that optimal allocation of computational resources between epistemic and aleatory calculations by means of convergence studies ensures accuracy within a limited budget. - Highlights: • Accident simulations considering uncertainties require intensive computations. • Tradeoff between accuracy and accident simulations is a challenge. • Optimal allocation between epistemic & aleatory computations ensures the tradeoff. • Online convergence gives an early indication of computational requirements. • Uncertainty propagation in DDET is examined on a tank problem solved analytically.

Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix

International Nuclear Information System (INIS)

Yamamoto, A.; Yasue, Y.; Endo, T.; Kodama, Y.; Ohoka, Y.; Tatsumi, M.

2012-01-01

An uncertainty estimation method for core safety parameters, for which measurement values are not obtained, is proposed. We empirically recognize the correlations among the prediction errors among core safety parameters, e.g., a correlation between the control rod worth and assembly relative power of corresponding position. Correlations of uncertainties among core safety parameters are theoretically estimated using the covariance of cross sections and sensitivity coefficients for core parameters. The estimated correlations among core safety parameters are verified through the direct Monte-Carlo sampling method. Once the correlation of uncertainties among core safety parameters is known, we can estimate the uncertainty of a safety parameter for which measurement value is not obtained. Furthermore, the correlations can be also used for the reduction of uncertainties of core safety parameters. (authors)

Assessment of uncertainties in severe accident management strategies

International Nuclear Information System (INIS)

Kastenberg, W.E.; Apostolakis, G.; Catton, I.; Dhir, V.K.; Okrent, D.

1990-01-01

Recent progress on the development of Probabilistic Risk Assessment (PRA) as a tool for qualifying nuclear reactor safety and on research devoted to severe accident phenomena has made severe accident management an achievable goal. Severe accident management strategies may involve operational changes, modification and/or addition of hardware, and institutional changes. In order to achieve the goal of managing severe accidents, a method for assessment of strategies must be developed which integrates PRA methodology and our current knowledge concerning severe accident phenomena, including uncertainty. The research project presented in this paper is aimed at delineating uncertainties in severe accident progression and their impact on severe accident management strategies

Assessment of uncertainty in full core reactor physics calculations using statistical methods

International Nuclear Information System (INIS)

McEwan, C.

2012-01-01

The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

Assessment of uncertainty in full core reactor physics calculations using statistical methods

Energy Technology Data Exchange (ETDEWEB)

McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

2012-07-01

The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

Subspace-based Inverse Uncertainty Quantification for Nuclear Data Assessment

Energy Technology Data Exchange (ETDEWEB)

Khuwaileh, B.A., E-mail: bakhuwai@ncsu.edu; Abdel-Khalik, H.S.

2015-01-15

Safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. An inverse problem can be defined and solved to assess the sources of uncertainty, and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this work a subspace-based algorithm for inverse sensitivity/uncertainty quantification (IS/UQ) has been developed to enable analysts account for all sources of nuclear data uncertainties in support of target accuracy assessment-type analysis. An approximate analytical solution of the optimization problem is used to guide the search for the dominant uncertainty subspace. By limiting the search to a subspace, the degrees of freedom available for the optimization search are significantly reduced. A quarter PWR fuel assembly is modeled and the accuracy of the multiplication factor and the fission reaction rate are used as reactor attributes whose uncertainties are to be reduced. Numerical experiments are used to demonstrate the computational efficiency of the proposed algorithm. Our ongoing work is focusing on extending the proposed algorithm to account for various forms of feedback, e.g., thermal-hydraulics and depletion effects.

Uncertainties in risk assessment and decision making

International Nuclear Information System (INIS)

Starzec, Peter; Purucker, Tom; Stewart, Robert

2008-02-01

The general concept for risk assessment in accordance with the Swedish model for contaminated soil implies that the toxicological reference value for a given receptor is first back-calculated to a corresponding concentration of a compound in soil and (if applicable) then modified with respect to e.g. background levels, acute toxicity, and factor of safety. This result in a guideline value that is subsequently compared to the observed concentration levels. Many sources of uncertainty exist when assessing whether the risk for a receptor is significant or not. In this study, the uncertainty aspects have been addressed from three standpoints: 1. Uncertainty in the comparison between the level of contamination (source) and a given risk criterion (e.g. a guideline value) and possible implications on subsequent decisions. This type of uncertainty is considered to be most important in situations where a contaminant is expected to be spatially heterogeneous without any tendency to form isolated clusters (hotspots) that can be easily delineated, i.e. where mean values are appropriate to compare to the risk criterion. 2. Uncertainty in spatial distribution of a contaminant. Spatial uncertainty should be accounted for when hotspots are to be delineated and the volume of soil contaminated with levels above a stated decision criterion has to be assessed (quantified). 3. Uncertainty in an ecological exposure model with regard to the moving pattern of a receptor in relation to spatial distribution of contaminant in question. The study points out that the choice of methodology to characterize the relation between contaminant concentration and a pre-defined risk criterion is governed by a conceptual perception of the contaminant's spatial distribution and also depends on the structure of collected data (observations). How uncertainty in transition from contaminant concentration into risk criterion can be quantified was demonstrated by applying hypothesis tests and the concept of

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

International Nuclear Information System (INIS)

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

1997-01-01

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

OECD/NEA expert group on uncertainty analysis for criticality safety assessment: Results of benchmark on sensitivity calculation (phase III)

Energy Technology Data Exchange (ETDEWEB)

Ivanova, T.; Laville, C. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay aux Roses (France); Dyrda, J. [Atomic Weapons Establishment AWE, Aldermaston, Reading, RG7 4PR (United Kingdom); Mennerdahl, D. [E Mennerdahl Systems EMS, Starvaegen 12, 18357 Taeby (Sweden); Golovko, Y.; Raskach, K.; Tsiboulia, A. [Inst. for Physics and Power Engineering IPPE, 1, Bondarenko sq., 249033 Obninsk (Russian Federation); Lee, G. S.; Woo, S. W. [Korea Inst. of Nuclear Safety KINS, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Bidaud, A.; Sabouri, P. [Laboratoire de Physique Subatomique et de Cosmologie LPSC, CNRS-IN2P3/UJF/INPG, Grenoble (France); Patel, A. [U.S. Nuclear Regulatory Commission (NRC), Washington, DC 20555-0001 (United States); Bledsoe, K.; Rearden, B. [Oak Ridge National Laboratory ORNL, M.S. 6170, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Gulliford, J.; Michel-Sendis, F. [OECD/NEA, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France)

2012-07-01

The sensitivities of the k{sub eff} eigenvalue to neutron cross sections have become commonly used in similarity studies and as part of the validation algorithm for criticality safety assessments. To test calculations of the sensitivity coefficients, a benchmark study (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (Expert Group on Uncertainty Analysis for Criticality Safety Assessment). This paper presents some sensitivity results generated by the benchmark participants using various computational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. The study demonstrates the performance of the tools. It also illustrates how model simplifications impact the sensitivity results and demonstrates the importance of 'implicit' (self-shielding) sensitivities. This work has been a useful step towards verification of the existing and developed sensitivity analysis methods. (authors)

Feedback from uncertainties propagation research projects conducted in different hydraulic fields: outcomes for engineering projects and nuclear safety assessment.

Science.gov (United States)

Bacchi, Vito; Duluc, Claire-Marie; Bertrand, Nathalie; Bardet, Lise

2017-04-01

In recent years, in the context of hydraulic risk assessment, much effort has been put into the development of sophisticated numerical model systems able reproducing surface flow field. These numerical models are based on a deterministic approach and the results are presented in terms of measurable quantities (water depths, flow velocities, etc…). However, the modelling of surface flows involves numerous uncertainties associated both to the numerical structure of the model, to the knowledge of the physical parameters which force the system and to the randomness inherent to natural phenomena. As a consequence, dealing with uncertainties can be a difficult task for both modelers and decision-makers [Ioss, 2011]. In the context of nuclear safety, IRSN assesses studies conducted by operators for different reference flood situations (local rain, small or large watershed flooding, sea levels, etc…), that are defined in the guide ASN N°13 [ASN, 2013]. The guide provides some recommendations to deal with uncertainties, by proposing a specific conservative approach to cover hydraulic modelling uncertainties. Depending of the situation, the influencing parameter might be the Strickler coefficient, levee behavior, simplified topographic assumptions, etc. Obviously, identifying the most influencing parameter and giving it a penalizing value is challenging and usually questionable. In this context, IRSN conducted cooperative (Compagnie Nationale du Rhone, I-CiTy laboratory of Polytech'Nice, Atomic Energy Commission, Bureau de Recherches Géologiques et Minières) research activities since 2011 in order to investigate feasibility and benefits of Uncertainties Analysis (UA) and Global Sensitivity Analysis (GSA) when applied to hydraulic modelling. A specific methodology was tested by using the computational environment Promethee, developed by IRSN, which allows carrying out uncertainties propagation study. This methodology was applied with various numerical models and in

Uncertainty and conservatism in safety evaluations based on a BEPU approach

International Nuclear Information System (INIS)

Yamaguchi, A.; Mizokami, S.; Kudo, Y.; Hotta, A.

2009-01-01

Atomic Energy Society of Japan has published 'Standard Method for Safety Evaluation using Best Estimate Code Based on Uncertainty and Scaling Analyses with Statistical Approach' to be applied to accidents and AOOs in the safety evaluation of LWRs. In this method, hereafter named as the AESJ-SSE (Statistical Safety Evaluation) method, identification and quantification of uncertainties will be performed and then a combination of the best estimate code and the evaluation of uncertainty propagation will be performed. Uncertainties are categorized into bias and variability. In general, bias is related to our state-of-knowledge on uncertainty objects (modeling, scaling, input data, etc.) while variability reflects stochastic features involved in these objects. Considering many kinds of uncertainties in thermal-hydraulics models and experimental databases show variabilities that will be strongly influenced by our state of knowledge, it seems reasonable that these variabilities are also related to state-of-knowledge. The design basis events (DBEs) that are employed for licensing analyses form a main part of the given or prior conservatism. The regulatory acceptance criterion is also regarded as the prior conservatism. In addition to these prior conservatisms, a certain amount of the posterior conservatism is added with maintaining intimate relationships with state-of-knowledge. In the AESJ-SSE method, this posterior conservatism can be incorporated into the safety evaluation in a combination of the following three ways, (1) broadening ranges of variability relevant to uncertainty objects, (2) employing more disadvantageous biases relevant to uncertainty objects and (3) adding an extra bias to the safety evaluation results. Knowing implemented quantitative bases of uncertainties and conservatism, the AESJ-SSE method provides a useful ground for rational decision-making. In order to seek for 'the best estimation' as well as reasonably setting the analytical margin, a degree

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

NARCIS (Netherlands)

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

1995-01-01

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

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

OpenAIRE

Nganje, William E.

1999-01-01

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

Implicit Treatment of Technical Specification and Thermal Hydraulic Parameter Uncertainties in Gaussian Process Model to Estimate Safety Margin

Directory of Open Access Journals (Sweden)

Douglas A. Fynan

2016-06-01

Full Text Available The Gaussian process model (GPM is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU and Level 1 probabilistic safety assessment (PSA success criteria definitions while dealing with a large number of uncertainties.

Uncertainty analysis with a view towards applications in accident consequence assessments

International Nuclear Information System (INIS)

Fischer, F.; Erhardt, J.

1985-09-01

Since the publication of the US-Reactor Safety Study WASH-1400 there has been an increasing interest to develop and apply methods which allow to quantify the uncertainty inherent in probabilistic risk assessments (PRAs) and accident consequence assessments (ACAs) for installations of the nuclear fuel cycle. Research and development in this area is forced by the fact that PRA and ACA are more and more used for comparative, decisive and fact finding studies initiated by industry and regulatory commissions. This report summarizes and reviews some of the main methods and gives some hints to do sensitivity and uncertainty analyses. Some first investigations aiming at the application of the method mentioned above to a submodel of the ACA-code UFOMOD (KfK) are presented. Sensitivity analyses and some uncertainty studies an important submodel of UFOMOD are carried out to identify the relevant parameters for subsequent uncertainty calculations. (orig./HP) [de

A review of the uncertainties in the assessment of radiological consequences of spent nuclear fuel disposal

International Nuclear Information System (INIS)

Wiborgh, M.; Elert, M.; Hoeglund, L.O.; Jones, C.; Grundfelt, B.; Skagius, K.; Bengtsson, A.

1992-06-01

Radioactive waste disposal systems for spent nuclear fuel are designed to isolate the radioactive waste from the human environment for long period of time. The isolation is provided by a combination of engineered and natural barriers. Safety assessments are performed to describe and quantify the performance of the individual barriers and the disposal system over long-term periods. These assessments will always be associated with uncertainties. Uncertainties can originate from the variability of natural systems and will also be introduced in the predictive modelling performed to quantitatively evaluate the behaviour of the disposal system as a consequence of the incomplete knowledge about the governing processes. Uncertainties in safety assessments can partly be reduced by additional measurements and research. The aim of this study has been to identify uncertainties in assessments of radiological consequences from the disposal of spent nuclear fuel based on the Swedish KBS-3 concept. The identified uncertainties have been classified with respect to their origin, i.e. in conceptual, modelling and data uncertainties. The possibilities to reduce the uncertainties are also commented upon. In assessments it is important to decrease uncertainties which are of major importance for the performance of the disposal system. These could to some extent be identified by uncertainty analysis. However, conceptual uncertainties and some type of model uncertainties are difficult to evaluate. To be able to decrease uncertainties in conceptual models, it is essential that the processes describing and influencing the radionuclide transport in the engineered and natural barriers are sufficiently understood. In this study a qualitative approach has been used. The importance of different barriers and processes are indicated by their influence on the release of some representative radionuclides. (122 refs.) (au)

Some concepts of model uncertainty for performance assessments of nuclear waste repositories

International Nuclear Information System (INIS)

Eisenberg, N.A.; Sagar, B.; Wittmeyer, G.W.

1994-01-01

Models of the performance of nuclear waste repositories will be central to making regulatory decisions regarding the safety of such facilities. The conceptual model of repository performance is represented by mathematical relationships, which are usually implemented as one or more computer codes. A geologic system may allow many conceptual models, which are consistent with the observations. These conceptual models may or may not have the same mathematical representation. Experiences in modeling the performance of a waste repository representation. Experiences in modeling the performance of a waste repository (which is, in part, a geologic system), show that this non-uniqueness of conceptual models is a significant source of model uncertainty. At the same time, each conceptual model has its own set of parameters and usually, it is not be possible to completely separate model uncertainty from parameter uncertainty for the repository system. Issues related to the origin of model uncertainty, its relation to parameter uncertainty, and its incorporation in safety assessments are discussed from a broad regulatory perspective. An extended example in which these issues are explored numerically is also provided

A Framework for Understanding Uncertainty in Seismic Risk Assessment.

Science.gov (United States)

Foulser-Piggott, Roxane; Bowman, Gary; Hughes, Martin

2017-10-11

A better understanding of the uncertainty that exists in models used for seismic risk assessment is critical to improving risk-based decisions pertaining to earthquake safety. Current models estimating the probability of collapse of a building do not consider comprehensively the nature and impact of uncertainty. This article presents a model framework to enhance seismic risk assessment and thus gives decisionmakers a fuller understanding of the nature and limitations of the estimates. This can help ensure that risks are not over- or underestimated and the value of acquiring accurate data is appreciated fully. The methodology presented provides a novel treatment of uncertainties in input variables, their propagation through the model, and their effect on the results. The study presents ranges of possible annual collapse probabilities for different case studies on buildings in different parts of the world, exposed to different levels of seismicity, and with different vulnerabilities. A global sensitivity analysis was conducted to determine the significance of uncertain variables. Two key outcomes are (1) that the uncertainty in ground-motion conversion equations has the largest effect on the uncertainty in the calculation of annual collapse probability; and (2) the vulnerability of a building appears to have an effect on the range of annual collapse probabilities produced, i.e., the level of uncertainty in the estimate of annual collapse probability, with less vulnerable buildings having a smaller uncertainty. © 2017 Society for Risk Analysis.

Sensitivity, uncertainty, and importance analysis of a risk assessment

International Nuclear Information System (INIS)

Andsten, R.S.; Vaurio, J.K.

1992-01-01

In this paper a number of supplementary studies and applications associated with probabilistic safety assessment (PSA) are described, including sensitivity and importance evaluations of failures, errors, systems, and groups of components. The main purpose is to illustrate the usefulness of a PSA for making decisions about safety improvements, training, allowed outage times, and test intervals. A useful measure of uncertainty importance is presented, and it points out areas needing development, such as reactor vessel aging phenomena, for reducing overall uncertainty. A time-dependent core damage frequency is also presented, illustrating the impact of testing scenarios and intervals. Tea methods and applications presented are based on the Level 1 PSA carried out for the internal initiating event of the Loviisa 1 nuclear power station. Steam generator leakages and associated operator actions are major contributors to the current core-damage frequency estimate of 2 x10 -4 /yr. The results are used to improve the plant and procedures and to guide future improvements

Uncertainty characteristics of EPA's ground-water transport model for low-level waste performance assessment

International Nuclear Information System (INIS)

Yim, Man-Sung

1995-01-01

Performance assessment is an essential step either in design or in licensing processes to ensure the safety of any proposed radioactive waste disposal facilities. Since performance assessment requires the use of computer codes, understanding the characteristics of computer models used and the uncertainties of the estimated results is important. The PRESTO-EPA code, which was the basis of the Environmental Protection Agency's analysis for low-level-waste rulemaking, is widely used for various performance assessment activities in the country with no adequate information available for the uncertainty characteristics of the results. In this study, the groundwater transport model PRESTO-EPA was examined based on the analysis of 14 C transport along with the investigation of uncertainty characteristics

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

International Nuclear Information System (INIS)

Dusic, Milorad; )

2013-01-01

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

Independent assessment for new nuclear reactor safety

Directory of Open Access Journals (Sweden)

D'Auria Francesco

2017-01-01

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

Independent assessment for new nuclear reactor safety

International Nuclear Information System (INIS)

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

2017-01-01

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

Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix

International Nuclear Information System (INIS)

Yamamoto, Akio; Yasue, Yoshihiro; Endo, Tomohiro; Kodama, Yasuhiro; Ohoka, Yasunori; Tatsumi, Masahiro

2013-01-01

An uncertainty reduction method for core safety parameters, for which measurement values are not obtained, is proposed. We empirically recognize that there exist some correlations among the prediction errors of core safety parameters, e.g., a correlation between the control rod worth and the assembly relative power at corresponding position. Correlations of errors among core safety parameters are theoretically estimated using the covariance of cross sections and sensitivity coefficients of core parameters. The estimated correlations of errors among core safety parameters are verified through the direct Monte Carlo sampling method. Once the correlation of errors among core safety parameters is known, we can estimate the uncertainty of a safety parameter for which measurement value is not obtained. (author)

Integrated Deterministic-Probabilistic Safety Assessment Methodologies

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-01

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

Uncertainty and sensitivity analysis in a Probabilistic Safety Analysis level-1

International Nuclear Information System (INIS)

Nunez Mc Leod, Jorge E.; Rivera, Selva S.

1996-01-01

A methodology for sensitivity and uncertainty analysis, applicable to a Probabilistic Safety Assessment Level I has been presented. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and systems response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well as different graphical visualization for the control of the study. (author)

Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

International Nuclear Information System (INIS)

Borges, Ronaldo Celem

2001-10-01

This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

MAPLE research reactor safety uncertainty assessment methodology

International Nuclear Information System (INIS)

Sills, H.E.; Duffey, R.B.; Andres, T.H.

1999-01-01

The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

Quantifying reactor safety margins: Application of CSAU [Code Scalability, Applicability and Uncertainty] methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

International Nuclear Information System (INIS)

Wulff, W.; Boyack, B.E.; Duffey, R.B.

1988-01-01

Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

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

Assessing Risk-Based Performance Indicators in Safety-Critical Systems for Nuclear Power Plants

OpenAIRE

TONT Gabriela

2011-01-01

The paper proposes framework for a multidisciplinary nuclear risk and safety assessment by modeling uncertainty and combining diverse evidence provided in such a way that it could be used to represent an entire argument about a system's dependability. The identified safety issues are being treated by means of probabilistic safety assessment (PSA). The behavior simulation of power plant in thepresence of risk factors is analyzed from the vulnerability, risk and functional safety viewpoints, hi...

Human factors in safety assessment. Safety culture assessment

International Nuclear Information System (INIS)

Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

1996-01-01

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

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

International Nuclear Information System (INIS)

2003-01-01

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

Identification of the Uncertainties for the Calibration of the Partial Safety Factors for Load in Tidal Turbines

Directory of Open Access Journals (Sweden)

Gaizka Zarraonandia Simeón

2016-03-01

Full Text Available Tidal energy is nowadays one of the fastest growing types of marine renewable energy. In particular, Horizontal Axis Tidal Turbines (HATTs are the most advanced designs and the most appropriate for standardization. This paper presents a review of actual design criteria focusing on the identification of the uncertainties that technology developers need to address during the design process. Key environmental parameters like turbine inflow conditions or predictions of extreme values are still grey areas due to the lack of site measurements and the uncertainty in metocean model predictions. A comparison of turbulence intensity characterization using different tools and at different points in time shows the uncertainty in the prediction of this parameter. Numerical models of HATTs are still quite uncertain, often dependent on experience of the people running them. In the reliability-based calibration of partial safety factors, the uncertainties need to be reflected on the limit state formulation. This paper analyses the different types of uncertainties present in the limit state equation. These uncertainties are assessed in terms of stochastic variables in the limit state equation. In some cases, advantage can be taken from the experience from offshore wind and oil and gas industries. Tidal turbines have a mixture of the uncertainties present in both industries with regard to partial safety factor calibration.

Biosphere modeling for safety assessment to high-level radioactive waste geological disposal. Application of reference biosphere methodology to safety assesment of geological disposal

International Nuclear Information System (INIS)

Baba, Tomoko; Ishihara, Yoshinao; Ishiguro, Katsuhiko; Suzuki, Yuji; Naito, Morimasa

2000-01-01

In the safety assessment of a high-level radioactive waste disposal system, it is required to estimate future radiological impacts on human beings. Consideration of living habits and the human environment in the future involves a large degree of uncertainty. To avoid endless speculation aimed at reducing such uncertainty, an approach is applied for identifying and justifying a 'reference biosphere' for use in safety assessment in Japan. considering a wide range of Japanese geological environments, saline specific reference biospheres' were developed using an approach consistent with the BIOMOVS II reference biosphere methodology. (author)

Overview of the ISAM safety assessment methodology

International Nuclear Information System (INIS)

Simeonov, G.

2003-01-01

The ISAM safety assessment methodology consists of the following key components: specification of the assessment context description of the disposal system development and justification of scenarios formulation and implementation of models running of computer codes and analysis and presentation of results. Common issues run through two or more of these assessment components, including: use of methodological and computer tools, collation and use of data, need to address various sources of uncertainty, building of confidence in the individual components, as well as the overall assessment. The importance of the iterative nature of the assessment should be recognised

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

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrik (comp.)

2006-09-15

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

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

International Nuclear Information System (INIS)

Sellin, Patrik

2006-09-01

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

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

International Nuclear Information System (INIS)

Pepin, G.; Sallaberry, C.

2003-01-01

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

Uncertainty quantification in flood risk assessment

Science.gov (United States)

Blöschl, Günter; Hall, Julia; Kiss, Andrea; Parajka, Juraj; Perdigão, Rui A. P.; Rogger, Magdalena; Salinas, José Luis; Viglione, Alberto

2017-04-01

Uncertainty is inherent to flood risk assessments because of the complexity of the human-water system, which is characterised by nonlinearities and interdependencies, because of limited knowledge about system properties and because of cognitive biases in human perception and decision-making. On top of the uncertainty associated with the assessment of the existing risk to extreme events, additional uncertainty arises because of temporal changes in the system due to climate change, modifications of the environment, population growth and the associated increase in assets. Novel risk assessment concepts are needed that take into account all these sources of uncertainty. They should be based on the understanding of how flood extremes are generated and how they change over time. They should also account for the dynamics of risk perception of decision makers and population in the floodplains. In this talk we discuss these novel risk assessment concepts through examples from Flood Frequency Hydrology, Socio-Hydrology and Predictions Under Change. We believe that uncertainty quantification in flood risk assessment should lead to a robust approach of integrated flood risk management aiming at enhancing resilience rather than searching for optimal defense strategies.

Fracture mechanics characteristics and associated safety margins for integrity assessment; Bruchmechanische Kennwerte und zugeordnete Sicherheitsfaktoren bei Integritaetsanalysen

Energy Technology Data Exchange (ETDEWEB)

Roos, E.; Schuler, X.; Stumpfrock, L.; Silcher, H. [Stuttgart Univ. (DE). Materialpruefungsanstalt (MPA)

2008-07-01

Within the integrity assessment of components and structural members of plants safety margins have to be applied, whose magnitude depend on several factors. Important factors influencing the magnitude of the safety margins are as for instance: Material behaviour (ductile / brittle behaviour), the event to be considered (local deformation / fracture), possible consequences of failure (human health, environmental damage, economic consequences) and many others. One important factor also is the fact, how precisely and reliably the appropriate material characteristics can be determined and how precisely and reliably the components behaviour can be predicted and assessed by means of this material characteristic. In contemporary safety assessment procedures by means of fracture mechanics evaluation tools (e.g. [1]) a concept of partial safety margins is proposed for application. The basic idea with this procedure is that only those sources of uncertainty have to be considered, which are relevant or may be relevant for the structure to be considered. For this purpose each source of possible uncertainty has to be quantified individually, finally only those singular safety margins are superimposed to a total safety margin which are relevant. The more the uncertainties have to be taken into account, the total safety margin to be applied, consequently will be larger. If some sources of uncertainty can be eliminated totally or can be minimized (for instance by a more reliable calculational procedure of the component loading or by more precise material characteristics), the total safety margin can be reduced. In this contribution the different procedures for the definition of safety margins within the integrity assessment by means of fracture mechanics procedures will be discussed. (orig.)

Validation of Fuel Performance Uncertainty for RIA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Nam-Gyu; Yoo, Jong-Sung; Jung, Yil-Sup [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

2016-10-15

To achieve this the computer code performance has to be validated based on the experimental results. And for the uncertainty quantification, important uncertainty parameters need to be selected, and combined uncertainty has to be evaluated with an acceptable statistical treatment. And important uncertainty parameters to the rod performance such as fuel enthalpy, fission gas release, cladding hoop strain etc. were chosen through the rigorous sensitivity studies. And their validity has been assessed by utilizing the experimental results, which were tested in CABRI and NSRR. Analysis results revealed that several tested rods were not bounded within combined fuel performance uncertainty. Assessment of fuel performance with an extended fuel power uncertainty on tested rods in NSRR and CABRI has been done. Analysis results showed that several tested rods were not bounded within calculated fuel performance uncertainty. This implies that the currently considered uncertainty range of the parameters is not enough to cover the fuel performance sufficiently.

Applications of Probabilistic Consequence Assessment Uncertainty Analysis for Plant Management (invited paper)

International Nuclear Information System (INIS)

Boardman, J.; Pearce, K.I.; Ponting, A.C.

2000-01-01

Probabilistic Consequence Assessment (PCA) models describe the dispersion of released radioactive materials and predict the resulting interaction with and influence on the environment and man. Increasing use is being made of PCA tools as an input to the evaluation and improvement of safety for nuclear installations. The nature and extent of the assessment performed varies considerably according to its intended purpose. Nevertheless with the increasing use of such techniques, greater attention has been given to the reliability of the methods used and the inherent uncertainty associated with their predictions. Uncertainty analyses can provide the decision-maker with information to quantify how uncertain the answer is and what drives that uncertainty. They often force a review of the baseline assumptions for any PCA methodology and provide a benchmark against which the impact of further changes in models and recommendations can be compared. This process provides valuable management information to help prioritise further actions or research. (author)

The role of quantitative uncertainty in the safety analysis of flammable gas accidents in Hanford waste tanks

International Nuclear Information System (INIS)

Bratzel, D.R.

1998-01-01

which will solve the two basic difficulties of defining the bounding case and assessing the impact of controls. The refined safety analysis does this by explicitly quantifying the effects of the uncertainty in the state of knowledge about accident phenomena and data and providing a consistent basis for calculating the impact of alternative control strategies on parameters that affect accident risk. The refined analysis allows the assessment of the risk impact of the variability in conditions (e.g., waste inventory) among storage tanks in the TWRS. Finally, the refined flammable gas accident safety analysis supports sensitivity studies to examine the impact on the results of differences in flammable gas accident perspectives

Uncertainty in ecological risk assessment: A statistician's view

International Nuclear Information System (INIS)

Smith, E.P.

1995-01-01

Uncertainty is a topic that has different meanings to researchers, modelers, managers and policy makers. The perspective of this presentation will be on the modeling view of uncertainty and its quantitative assessment. The goal is to provide some insight into how a statistician visualizes and addresses the issue of uncertainty in ecological risk assessment problems. In ecological risk assessment, uncertainty arises from many sources and is of different type depending on what is studies, where it is studied and how it is studied. Some major sources and their impact are described. A variety of quantitative approaches to modeling uncertainty are characterized and a general taxonomy given. Examples of risk assessments of lake acidification, power plant impact assessment and the setting of standards for chemicals will be used discuss approaches to quantitative assessment of uncertainty and some of the potential difficulties

Advanced methods for the risk, vulnerability and resilience assessment of safety-critical engineering components, systems and infrastructures, in the presence of uncertainties

International Nuclear Information System (INIS)

Pedroni, Nicolas

2016-01-01

Safety-critical industrial installations (e.g., nuclear plants) and infrastructures (e.g., power transmission networks) are complex systems composed by a multitude and variety of heterogeneous 'elements', which are highly interconnected and mutually dependent. In addition, such systems are affected by large uncertainties in the characterization of the failure and recovery behavior of their components, interconnections and interactions. Such characteristics raise concerns with respect to the system risk, vulnerability and resilience properties, which have to be accurately and precisely assessed for decision making purposes. In general, this entails the following main steps: (1) representation of the system to capture its main features; (2) construction of a mathematical model of the system; (3) simulation of the behavior of the system under various uncertain conditions to evaluate the relevant risk, vulnerability and resilience metrics by propagating the uncertainties through the mathematical model; (4) decision making to (optimally) determine the set of protective actions to effectively reduce (resp., increase) the system risk and vulnerability (resp., resilience). New methods to address these issues have been developed in this dissertation. Specifically, the research works have been carried out along two main axes: (1) the study of approaches for uncertainty modeling and quantification; (2) the development of advanced computational methods for the efficient system modeling, simulation and analysis in the presence of uncertainties. (author)

Hybrid probabilistic and possibilistic safety assessment. Methodology and application

International Nuclear Information System (INIS)

Kato, Kazuyuki; Amano, Osamu; Ueda, Hiroyoshi; Ikeda, Takao; Yoshida, Hideji; Takase, Hiroyasu

2002-01-01

This paper presents a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to the safety assessment of geological disposal of high-level radioactive waste. Uncertainties associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts, while variability was formulated by means of probability density functions (pdfs) based on available data sets. The exercise demonstrated the applicability of the new methodology and, in particular, its advantage in quantifying uncertainties based on expert opinion and in providing information on the dependence of assessment results on the level of conservatism. In addition, it was shown that sensitivity analysis can identify key parameters contributing to uncertainties associated with results of the overall assessment. The information mentioned above can be utilized to support decision-making and to guide the process of disposal system development and optimization of protection against potential exposure. (author)

Uncertainties in radioecological assessment models

International Nuclear Information System (INIS)

Hoffman, F.O.; Miller, C.W.; Ng, Y.C.

1983-01-01

Environmental radiological assessments rely heavily on the use of mathematical models. The predictions of these models are inherently uncertain because models are inexact representations of real systems. The major sources of this uncertainty are related to bias in model formulation and imprecision in parameter estimation. The magnitude of uncertainty is a function of the questions asked of the model and the specific radionuclides and exposure pathways of dominant importance. It is concluded that models developed as research tools should be distinguished from models developed for assessment applications. Furthermore, increased model complexity does not necessarily guarantee increased accuracy. To improve the realism of assessment modeling, stochastic procedures are recommended that translate uncertain parameter estimates into a distribution of predicted values. These procedures also permit the importance of model parameters to be ranked according to their relative contribution to the overall predicted uncertainty. Although confidence in model predictions can be improved through site-specific parameter estimation and increased model validation, health risk factors and internal dosimetry models will probably remain important contributors to the amount of uncertainty that is irreducible. 41 references, 4 figures, 4 tables

A comparison of integrated safety analysis and probabilistic risk assessment

International Nuclear Information System (INIS)

Damon, Dennis R.; Mattern, Kevin S.

2013-01-01

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

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

Science.gov (United States)

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

2014-01-01

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

Quantifying remarks to the question of uncertainties of the 'general dose assessment fundamentals'

International Nuclear Information System (INIS)

Brenk, H.D.; Vogt, K.J.

1982-12-01

Dose prediction models are always subject to uncertainties due to a number of factors including deficiencies in the model structure and uncertainties of the model input parameter values. In lieu of validation experiments the evaluation of these uncertainties is restricted to scientific judgement. Several attempts have been made in the literature to evaluate the uncertainties of the current dose assessment models resulting from uncertainties of the model input parameter values using stochastic approaches. Less attention, however, has been paid to potential sources of systematic over- and underestimations of the predicted doses due to deficiencies in the model structure. The present study addresses this aspect with regard to dose assessment models currently used for regulatory purposes. The influence of a number of basic simplifications and conservative assumptions has been investigated. Our systematic approach is exemplified by a comparison of doses evaluated on the basis of the regulatory guide model and a more realistic model respectively. This is done for 3 critical exposure pathways. As a result of this comparison it can be concluded that the currently used regularoty-type models include significant safety factors resulting in a systematic overprediction of dose to man up to two orders of magnitude. For this reason there are some indications that these models usually more than compensate the bulk of the stochastic uncertainties caused by the variability of the input parameter values. (orig.) [de

Uncertainties and severe-accident management

International Nuclear Information System (INIS)

Kastenberg, W.E.

1991-01-01

Severe-accident management can be defined as the use of existing and or alternative resources, systems, and actions to prevent or mitigate a core-melt accident. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies

Overview of methods for uncertainty analysis and sensitivity analysis in probabilistic risk assessment

International Nuclear Information System (INIS)

Iman, R.L.; Helton, J.C.

1985-01-01

Probabilistic Risk Assessment (PRA) is playing an increasingly important role in the nuclear reactor regulatory process. The assessment of uncertainties associated with PRA results is widely recognized as an important part of the analysis process. One of the major criticisms of the Reactor Safety Study was that its representation of uncertainty was inadequate. The desire for the capability to treat uncertainties with the MELCOR risk code being developed at Sandia National Laboratories is indicative of the current interest in this topic. However, as yet, uncertainty analysis and sensitivity analysis in the context of PRA is a relatively immature field. In this paper, available methods for uncertainty analysis and sensitivity analysis in a PRA are reviewed. This review first treats methods for use with individual components of a PRA and then considers how these methods could be combined in the performance of a complete PRA. In the context of this paper, the goal of uncertainty analysis is to measure the imprecision in PRA outcomes of interest, and the goal of sensitivity analysis is to identify the major contributors to this imprecision. There are a number of areas that must be considered in uncertainty analysis and sensitivity analysis for a PRA: (1) information, (2) systems analysis, (3) thermal-hydraulic phenomena/fission product behavior, (4) health and economic consequences, and (5) display of results. Each of these areas and the synthesis of them into a complete PRA are discussed

Reactor pressure vessels safety and reliability - certainty and uncertainty

International Nuclear Information System (INIS)

O'Neil, R.

1977-01-01

In the paper, it is suggested that the hazard to the population which would result from vessel failure rate of the order of 10 -6 to 10 -7 per vessel year could be acceptable to society on the basis of other natural and man-made risks. The paper considers the problems of demonstrating safety by calculation based on fracture mechanics, and indicates some of the uncertainties, and inconsistencies in the theory, particularly the effect of cracks in locally degraded volumes of material. The phenomenon of crack arrest is considered, and attention is drawn to the uncertainties as indicated at least by some tests. There is need for speedy resolution of this problem. The uncertainties in material properties, heat treatment and residual stresses are considered, and a proposed upper limit for residual defects ('original sin') is proposed. (orig.) [de

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

Science.gov (United States)

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

2017-01-01

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

Treatment of uncertainty in low-level waste performance assessment

International Nuclear Information System (INIS)

Kozak, M.W.; Olague, N.E.; Gallegos, D.P.; Rao, R.R.

1991-01-01

Uncertainties arise from a number of different sources in low-level waste performance assessment. In this paper the types of uncertainty are reviewed, and existing methods for quantifying and reducing each type of uncertainty are discussed. These approaches are examined in the context of the current low-level radioactive waste regulatory performance objectives, which are deterministic. The types of uncertainty discussed in this paper are model uncertainty, uncertainty about future conditions, and parameter uncertainty. The advantages and disadvantages of available methods for addressing uncertainty in low-level waste performance assessment are presented. 25 refs

Uncertainties in Safety Analysis. A literature review

International Nuclear Information System (INIS)

Ekberg, C.

1995-05-01

The purpose of the presented work has been to give a short summary of the origins of many uncertainties arising in the designing and performance assessment of a repository for spent nuclear fuel. Some different methods to treat these uncertainties is also included. The methods and conclusions are in many cases general in the sense that they are applicable to many other disciplines where simulations are used. As a conclusion it may be noted that uncertainties of different origin have been discussed and debated, but one large group, e.g. computer simulations, where the methods to make a more explicit investigation exists, have not been investigated in a satisfying way. 50 refs

Uncertainties in Safety Analysis. A literature review

Energy Technology Data Exchange (ETDEWEB)

Ekberg, C [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry

1995-05-01

The purpose of the presented work has been to give a short summary of the origins of many uncertainties arising in the designing and performance assessment of a repository for spent nuclear fuel. Some different methods to treat these uncertainties is also included. The methods and conclusions are in many cases general in the sense that they are applicable to many other disciplines where simulations are used. As a conclusion it may be noted that uncertainties of different origin have been discussed and debated, but one large group, e.g. computer simulations, where the methods to make a more explicit investigation exists, have not been investigated in a satisfying way. 50 refs.

Risk Assessment Uncertainties in Cybersecurity Investments

Directory of Open Access Journals (Sweden)

Andrew Fielder

2018-06-01

Full Text Available When undertaking cybersecurity risk assessments, it is important to be able to assign numeric values to metrics to compute the final expected loss that represents the risk that an organization is exposed to due to cyber threats. Even if risk assessment is motivated by real-world observations and data, there is always a high chance of assigning inaccurate values due to different uncertainties involved (e.g., evolving threat landscape, human errors and the natural difficulty of quantifying risk. Existing models empower organizations to compute optimal cybersecurity strategies given their financial constraints, i.e., available cybersecurity budget. Further, a general game-theoretic model with uncertain payoffs (probability-distribution-valued payoffs shows that such uncertainty can be incorporated in the game-theoretic model by allowing payoffs to be random. This paper extends previous work in the field to tackle uncertainties in risk assessment that affect cybersecurity investments. The findings from simulated examples indicate that although uncertainties in cybersecurity risk assessment lead, on average, to different cybersecurity strategies, they do not play a significant role in the final expected loss of the organization when utilising a game-theoretic model and methodology to derive these strategies. The model determines robust defending strategies even when knowledge regarding risk assessment values is not accurate. As a result, it is possible to show that the cybersecurity investments’ tool is capable of providing effective decision support.

Reliability and safety engineering

CERN Document Server

Verma, Ajit Kumar; Karanki, Durga Rao

2016-01-01

Reliability and safety are core issues that must be addressed throughout the life cycle of engineering systems. Reliability and Safety Engineering presents an overview of the basic concepts, together with simple and practical illustrations. The authors present reliability terminology in various engineering fields, viz.,electronics engineering, software engineering, mechanical engineering, structural engineering and power systems engineering. The book describes the latest applications in the area of probabilistic safety assessment, such as technical specification optimization, risk monitoring and risk informed in-service inspection. Reliability and safety studies must, inevitably, deal with uncertainty, so the book includes uncertainty propagation methods: Monte Carlo simulation, fuzzy arithmetic, Dempster-Shafer theory and probability bounds. Reliability and Safety Engineering also highlights advances in system reliability and safety assessment including dynamic system modeling and uncertainty management. Cas...

Regulatory risk assessments: Is there a need to reduce uncertainty and enhance robustness?

Science.gov (United States)

Snodin, D J

2015-12-01

A critical evaluation of several recent regulatory risk assessments has been undertaken. These relate to propyl paraben (as a food additive, cosmetic ingredient or pharmaceutical excipient), cobalt (in terms of a safety-based limit for pharmaceuticals) and the cancer Threshold of Toxicological Concern as applied to food contaminants and pharmaceutical impurities. In all cases, a number of concerns can be raised regarding the reliability of the current assessments, some examples being absence of data audits, use of single-dose and/or non-good laboratory practice studies to determine safety metrics, use of a biased data set and questionable methodology and lack of consistency with precedents and regulatory guidance. Drawing on these findings, a set of recommendations is provided to reduce uncertainty and improve the quality and robustness of future regulatory risk assessments. © The Author(s) 2015.

Road safety performance measures and AADT uncertainty from short-term counts.

Science.gov (United States)

Milligan, Craig; Montufar, Jeannette; Regehr, Jonathan; Ghanney, Bartholomew

2016-12-01

The objective of this paper is to enable better risk analysis of road safety performance measures by creating the first knowledge base on uncertainty surrounding annual average daily traffic (AADT) estimates when the estimates are derived by expanding short-term counts with the individual permanent counter method. Many road safety performance measures and performance models use AADT as an input. While there is an awareness that the input suffers from uncertainty, the uncertainty is not well known or accounted for. The paper samples data from a set of 69 permanent automatic traffic recorders in Manitoba, Canada, to simulate almost 2 million short-term counts over a five year period. These short-term counts are expanded to AADT estimates by transferring temporal information from a directly linked nearby permanent count control station, and the resulting AADT values are compared to a known reference AADT to compute errors. The impacts of five factors on AADT error are considered: length of short-term count, number of short-term counts, use of weekday versus weekend counts, distance from a count to its expansion control station, and the AADT at the count site. The mean absolute transfer error for expanded AADT estimates is 6.7%, and this value varied by traffic pattern group from 5% to 10.5%. Reference percentiles of the error distribution show that almost all errors are between -20% and +30%. Error decreases substantially by using a 48-h count instead of a 24-h count, and only slightly by using two counts instead of one. Weekday counts are superior to weekend counts, especially if the count is only 24h. Mean absolute transfer error increases with distance to control station (elasticity 0.121, p=0.001), and increases with AADT (elasticity 0.857, proad safety performance measures that use AADT as inputs. Analytical frameworks for such analysis exist but are infrequently used in road safety because the evidence base on AADT uncertainty is not well developed. Copyright �

Regulatory status on the safety assessment of a HLW repository in other countries

International Nuclear Information System (INIS)

Lee, Sung Ho; Hwang, Yong Soo

2008-12-01

To construct a HLW repository, it is essential to meet the requirements on the regulation for a deep geological disposal. Even if the construction of a HLW repository is determined positively, technical standards which assert the performance of a repository will be needed. Among various technical standards, safety assessment based on the repository evolution in the future will play an important role in the licensing process. The foreign countries' technical standards on the safety assessment of a HLW repository may be an indicator to carry out the R and D activities on geological disposal effectively. In this report, assessment period, limit of radiation dose and uncertainty related to the safety assessment are investigated and analyzed in detail. Especially, the technical reviews of USA regulation bodies seems to be reasonable in the point of the intrinsic attribute of safety assessment

Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

International Nuclear Information System (INIS)

Ergun, E; Kadiroglu, O.S.

1999-01-01

The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

Dealing with uncertainties in environmental burden of disease assessment

Directory of Open Access Journals (Sweden)

van der Sluijs Jeroen P

2009-04-01

Full Text Available Abstract Disability Adjusted Life Years (DALYs combine the number of people affected by disease or mortality in a population and the duration and severity of their condition into one number. The environmental burden of disease is the number of DALYs that can be attributed to environmental factors. Environmental burden of disease estimates enable policy makers to evaluate, compare and prioritize dissimilar environmental health problems or interventions. These estimates often have various uncertainties and assumptions which are not always made explicit. Besides statistical uncertainty in input data and parameters – which is commonly addressed – a variety of other types of uncertainties may substantially influence the results of the assessment. We have reviewed how different types of uncertainties affect environmental burden of disease assessments, and we give suggestions as to how researchers could address these uncertainties. We propose the use of an uncertainty typology to identify and characterize uncertainties. Finally, we argue that uncertainties need to be identified, assessed, reported and interpreted in order for assessment results to adequately support decision making.

Statistically based uncertainty assessments in nuclear risk analysis

International Nuclear Information System (INIS)

Spencer, F.W.; Diegert, K.V.; Easterling, R.G.

1987-01-01

Over the last decade, the problems of estimation and uncertainty assessment in probabilistics risk assessment (PRAs) have been addressed in a variety of NRC and industry-sponsored projects. These problems have received attention because of a recognition that major uncertainties in risk estimation exist, which can be reduced by collecting more and better data and other information, and because of a recognition that better methods for assessing these uncertainties are needed. In particular, a clear understanding of the nature and magnitude of various sources of uncertainty is needed to facilitate descision-making on possible plant changes and research options. Recent PRAs have employed methods of probability propagation, sometimes involving the use of Bayes Theorem, and intended to formalize the use of ''engineering judgment'' or ''expert opinion.'' All sources, or feelings, of uncertainty are expressed probabilistically, so that uncertainty analysis becomes simply a matter of probability propagation. Alternatives to forcing a probabilistic framework at all stages of a PRA are a major concern in this paper, however

The use of safety indicators in the assessment of radioactive waste disposal

International Nuclear Information System (INIS)

Wingefors, S.; Westerlind, M.; Gera, F.

1999-01-01

The most widely used criteria for disposal are limits or constraints on individual dose or risk, and these have been introduced in most national legal frameworks. There is general agreement that future generations have the right to the same level of protection as the current generation. Even if quantitative criteria corresponding to the required level of protection can be (and have been) defined, it is a great challenge to demonstrate compliance with these criteria. The difficulties are to large extent due to the long time-scales needed to be considered in radioactive waste disposal. The future cannot be predicted in detail but instead different scenarios, with different probabilities of occurrence, must be assessed. Some parts of a disposal system can be predicted or analysed with high confidence for very long periods of time, e.g. geological formations, while for example the evolution of the biosphere, and in particular the society, become quite uncertain within less than one thousand years. Thus, there may be considerable uncertainty in doses (or risks) derived from the safety assessment of a repository. Due to these unavoidable uncertainties it is believed advantageous to use multiple approaches in the safety assessment and to identify different indicators for the repository safety ('multiple-lines-of-reasoning'). The most fundamental safety indicators are dose/risk but complementary indicators have been suggested, in particular flux and environmental concentration of radionuclides. This presentation is focussed on fluxes and concentrations as complementary safety indicators. Other safety indicators, e.g. transfer times, are mentioned only briefly

Critical loads - assessment of uncertainty

Energy Technology Data Exchange (ETDEWEB)

Barkman, A.

1998-10-01

The effects of data uncertainty in applications of the critical loads concept were investigated on different spatial resolutions in Sweden and northern Czech Republic. Critical loads of acidity (CL) were calculated for Sweden using the biogeochemical model PROFILE. Three methods with different structural complexity were used to estimate the adverse effects of S0{sub 2} concentrations in northern Czech Republic. Data uncertainties in the calculated critical loads/levels and exceedances (EX) were assessed using Monte Carlo simulations. Uncertainties within cumulative distribution functions (CDF) were aggregated by accounting for the overlap between site specific confidence intervals. Aggregation of data uncertainties within CDFs resulted in lower CL and higher EX best estimates in comparison with percentiles represented by individual sites. Data uncertainties were consequently found to advocate larger deposition reductions to achieve non-exceedance based on low critical loads estimates on 150 x 150 km resolution. Input data were found to impair the level of differentiation between geographical units at all investigated resolutions. Aggregation of data uncertainty within CDFs involved more constrained confidence intervals for a given percentile. Differentiation as well as identification of grid cells on 150 x 150 km resolution subjected to EX was generally improved. Calculation of the probability of EX was shown to preserve the possibility to differentiate between geographical units. Re-aggregation of the 95%-ile EX on 50 x 50 km resolution generally increased the confidence interval for each percentile. Significant relationships were found between forest decline and the three methods addressing risks induced by S0{sub 2} concentrations. Modifying S0{sub 2} concentrations by accounting for the length of the vegetation period was found to constitute the most useful trade-off between structural complexity, data availability and effects of data uncertainty. Data

Incorporating uncertainties into risk assessment with an application to the exploratory studies facilities at Yucca Mountain

International Nuclear Information System (INIS)

Fathauer, P.M.

1995-08-01

A methodology that incorporates variability and reducible sources of uncertainty into the probabilistic and consequence components of risk was developed. The method was applied to the north tunnel of the Exploratory Studies Facility at Yucca Mountain in Nevada. In this assessment, variability and reducible sources of uncertainty were characterized and propagated through the risk assessment models using a Monte Carlo based software package. The results were then manipulated into risk curves at the 5% and 95% confidence levels for both the variability and overall uncertainty analyses, thus distinguishing between variability and reducible sources of uncertainty. In the Yucca Mountain application, the designation of the north tunnel as an item important to public safety, as defined by 10 CFR 60, was determined. Specifically, the annual frequency of a rock fall breaching a waste package causing an off-site dose of 500 mrem (5x10 -3 Sv) was calculated. The annual frequency, taking variability into account, ranged from 1.9x10 -9 per year at the 5% confidence level to 2.5x10 -9 per year at the 95% confidence level. The frequency range after including all uncertainty was 9.5x10 -10 to 1.8x10 -8 per year. The maximum observable frequency, at the 100% confidence level, was 4.9x10 -8 per year. This is below the 10 -6 per year frequency criteria of 10 CFR 60. Therefore, based on this work, the north tunnel does not fall under the items important to public safety designation for the event studied

Assessing scenario and parametric uncertainties in risk analysis: a model uncertainty audit

International Nuclear Information System (INIS)

Tarantola, S.; Saltelli, A.; Draper, D.

1999-01-01

In the present study a process of model audit is addressed on a computational model used for predicting maximum radiological doses to humans in the field of nuclear waste disposal. Global uncertainty and sensitivity analyses are employed to assess output uncertainty and to quantify the contribution of parametric and scenario uncertainties to the model output. These tools are of fundamental importance for risk analysis and decision making purposes

Robust Trajectory Optimization of a Ski Jumper for Uncertainty Influence and Safety Quantification

Directory of Open Access Journals (Sweden)

Patrick Piprek

2018-02-01

Full Text Available This paper deals with the development of a robust optimal control framework for a previously developed multi-body ski jumper simulation model by the authors. This framework is used to model uncertainties acting on the jumper during his jump, e.g., wind or mass, to enhance the performance, but also to increase the fairness and safety of the competition. For the uncertainty modeling the method of generalized polynomial chaos together with the discrete expansion by stochastic collocation is applied: This methodology offers a very flexible framework to model multiple uncertainties using a small number of required optimizations to calculate an uncertain trajectory. The results are then compared to the results of the Latin-Hypercube sampling method to show the correctness of the applied methods. Finally, the results are examined with respect to two major metrics: First, the influence of the uncertainties on the jumper, his positioning with respect to the air, and his maximal achievable flight distance are examined. Then, the results are used in a further step to quantify the safety of the jumper.

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

Many advanced light water reactor designs incorporate passive rather than active safety features for front-line accident response. A method for evaluating the reliability of these passive systems in the context of probabilistic risk assessment has been developed at Sandia National Laboratories. This method addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. These processes provide the system's driving force; examples are natural circulation and gravity-induced injection. This paper describes the method, and provides some preliminary results of application of the approach to the Westinghouse AP600 design

Uncertainties in environmental impact assessments due to expert opinion. Case study. Radioactive waste in Slovenia

International Nuclear Information System (INIS)

Kontic, B.; Ravnik, M.

1998-01-01

A comprehensive study was done at the J. Stefan Institute in Ljubljana and the School of Environmental Sciences in Nova Gorica in relation to sources of uncertainties in long-term environmental impact assessment (EIA). Under the research two main components were examined: first, methodology of the preparation of an EIA, and second validity of an expert opinion. Following the findings of the research a survey was performed in relation to assessing acceptability of radioactive waste repository by the regulatory. The components of dose evaluation in different time frames were examined in terms of susceptibility to uncertainty. Uncertainty associated to human exposure in the far future is so large that dose and risk, as individual numerical indicators of safety, by our opinion, should not be used in compliance assessment for radioactive waste repository. On the other hand, results of the calculations on the amount and activity of low and intermediate level waste and the spent fuel from the Krsko NPP show that expert's understanding of the treated questions can be expressed in transparent way giving credible output of the models used.(author)

DECOVALEX III PROJECT. Thermal-Hydro-Mechanical Coupled Processes in Safety Assessments. Report of Task 4

International Nuclear Information System (INIS)

Andersson, Johan

2005-02-01

A part (Task 4) of the International DECOVALEX III project on coupled thermo-hydro-mechanical (T-H-M) processes focuses on T-H-M modelling applications in safety and performance assessment of deep geological nuclear waste repositories. A previous phase, DECOVALEX II, saw a need to improve such modelling. In order to address this need Task 4 of DECOVALEX III has: Analysed two major T-H-M experiments (Task 1 and Task 2) and three different Bench Mark Tests (Task 3) set-up to explore the significance of T-H-M in some potentially important safety assessment applications. Compiled and evaluated the use of T-H-M modelling in safety assessments at the time of the year 2000. Organised a forum a forum of interchange between PA-analysts and THM modelers at each DECOVALEX III workshop. Based on this information the current report discusses the findings and strives for reaching recommendations as regards good practices in addressing coupled T-H-M issues in safety assessments. The full development of T-H-M modelling is still at an early stage and it is not evident whether current codes provide the information that is required. However, although the geosphere is a system of fully coupled processes, this does not directly imply that all existing coupled mechanisms must be represented numerically. Modelling is conducted for specific purposes and the required confidence level should be considered. It is necessary to match the confidence level with the modelling objective. Coupled THM modelling has to incorporate uncertainties. These uncertainties mainly concern uncertainties in the conceptual model and uncertainty in data. Assessing data uncertainty is important when judging the need to model coupled processes. Often data uncertainty is more significant than the coupled effects. The emphasis on the need for THM modelling differs among disciplines. For geological radioactive waste disposal in crystalline and other similar hard rock formations DECOVALEX III shows it is essential to

DECOVALEX III PROJECT. Thermal-Hydro-Mechanical Coupled Processes in Safety Assessments. Report of Task 4

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan [JA Streamflow AB, Aelvsjoe (Sweden)

2005-02-15

A part (Task 4) of the International DECOVALEX III project on coupled thermo-hydro-mechanical (T-H-M) processes focuses on T-H-M modelling applications in safety and performance assessment of deep geological nuclear waste repositories. A previous phase, DECOVALEX II, saw a need to improve such modelling. In order to address this need Task 4 of DECOVALEX III has: Analysed two major T-H-M experiments (Task 1 and Task 2) and three different Bench Mark Tests (Task 3) set-up to explore the significance of T-H-M in some potentially important safety assessment applications. Compiled and evaluated the use of T-H-M modelling in safety assessments at the time of the year 2000. Organised a forum a forum of interchange between PA-analysts and THM modelers at each DECOVALEX III workshop. Based on this information the current report discusses the findings and strives for reaching recommendations as regards good practices in addressing coupled T-H-M issues in safety assessments. The full development of T-H-M modelling is still at an early stage and it is not evident whether current codes provide the information that is required. However, although the geosphere is a system of fully coupled processes, this does not directly imply that all existing coupled mechanisms must be represented numerically. Modelling is conducted for specific purposes and the required confidence level should be considered. It is necessary to match the confidence level with the modelling objective. Coupled THM modelling has to incorporate uncertainties. These uncertainties mainly concern uncertainties in the conceptual model and uncertainty in data. Assessing data uncertainty is important when judging the need to model coupled processes. Often data uncertainty is more significant than the coupled effects. The emphasis on the need for THM modelling differs among disciplines. For geological radioactive waste disposal in crystalline and other similar hard rock formations DECOVALEX III shows it is essential to

Fuzzy randomness uncertainty in civil engineering and computational mechanics

CERN Document Server

Möller, Bernd

2004-01-01

This book, for the first time, provides a coherent, overall concept for taking account of uncertainty in the analysis, the safety assessment, and the design of structures. The reader is introduced to the problem of uncertainty modeling and familiarized with particular uncertainty models. For simultaneously considering stochastic and non-stochastic uncertainty the superordinated uncertainty model fuzzy randomness, which contains real valued random variables as well as fuzzy variables as special cases, is presented. For this purpose basic mathematical knowledge concerning the fuzzy set theory and the theory of fuzzy random variables is imparted. The body of the book comprises the appropriate quantification of uncertain structural parameters, the fuzzy and fuzzy probabilistic structural analysis, the fuzzy probabilistic safety assessment, and the fuzzy cluster structural design. The completely new algorithms are described in detail and illustrated by way of demonstrative examples.

Safety assessments for deep geological disposal of radioactive wastes

International Nuclear Information System (INIS)

Lyon, R.B.

1984-01-01

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

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

International Nuclear Information System (INIS)

Ivanov, Evgeny; Ivanova, Tatiana; Pignet, Sophie

2013-01-01

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

Bedrock Kd data and uncertainty assessment for application in SR-Site geosphere transport calculations

International Nuclear Information System (INIS)

Crawford, James

2010-12-01

The safety assessment SR-Site is undertaken to assess the safety of a potential geologic repository for spent nuclear fuel at the Forsmark and Laxemar sites. The present report is one of several reports that form the data input to SR-Site and contains a compilation of recommended K d data (i.e. linear partitioning coefficients) for safety assessment modelling of geosphere radionuclide transport. The data are derived for rock types and groundwater compositions distinctive of the site investigation areas at Forsmark and Laxemar. Data have been derived for all elements and redox states considered of importance for far-field dose estimates as described in /SKB 2010d/. The K d data are given in the form of lognormal distributions characterised by a mean (μ) and standard deviation (σ). Upper and lower limits for the uncertainty range of the recommended data are defined by the 2.5% and 97.5% percentiles of the empirical data sets. The best estimate K d value for use in deterministic calculations is given as the median of the K d distribution

Where do uncertainties reside within environmental risk assessments? Expert opinion on uncertainty distributions for pesticide risks to surface water organisms.

Science.gov (United States)

Skinner, Daniel J C; Rocks, Sophie A; Pollard, Simon J T

2016-12-01

A reliable characterisation of uncertainties can aid uncertainty identification during environmental risk assessments (ERAs). However, typologies can be implemented inconsistently, causing uncertainties to go unidentified. We present an approach based on nine structured elicitations, in which subject-matter experts, for pesticide risks to surface water organisms, validate and assess three dimensions of uncertainty: its level (the severity of uncertainty, ranging from determinism to ignorance); nature (whether the uncertainty is epistemic or aleatory); and location (the data source or area in which the uncertainty arises). Risk characterisation contains the highest median levels of uncertainty, associated with estimating, aggregating and evaluating the magnitude of risks. Regarding the locations in which uncertainty is manifest, data uncertainty is dominant in problem formulation, exposure assessment and effects assessment. The comprehensive description of uncertainty described will enable risk analysts to prioritise the required phases, groups of tasks, or individual tasks within a risk analysis according to the highest levels of uncertainty, the potential for uncertainty to be reduced or quantified, or the types of location-based uncertainty, thus aiding uncertainty prioritisation during environmental risk assessments. In turn, it is expected to inform investment in uncertainty reduction or targeted risk management action. Copyright © 2016 The Authors. Published by Elsevier B.V. All rights reserved.

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

International Nuclear Information System (INIS)

Kim, Ji Young; Kang, Keum Seok

2011-01-01

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

Computational Methods for Sensitivity and Uncertainty Analysis in Criticality Safety

International Nuclear Information System (INIS)

Broadhead, B.L.; Childs, R.L.; Rearden, B.T.

1999-01-01

Interest in the sensitivity methods that were developed and widely used in the 1970s (the FORSS methodology at ORNL among others) has increased recently as a result of potential use in the area of criticality safety data validation procedures to define computational bias, uncertainties and area(s) of applicability. Functional forms of the resulting sensitivity coefficients can be used as formal parameters in the determination of applicability of benchmark experiments to their corresponding industrial application areas. In order for these techniques to be generally useful to the criticality safety practitioner, the procedures governing their use had to be updated and simplified. This paper will describe the resulting sensitivity analysis tools that have been generated for potential use by the criticality safety community

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.

Economic aspects of risk assessment in chemical safety

Energy Technology Data Exchange (ETDEWEB)

Drummond, M F; Shannon, H S

1986-05-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Uncertainty Assessment in Urban Storm Water Drainage Modelling

DEFF Research Database (Denmark)

Thorndahl, Søren

The object of this paper is to make an overall description of the author's PhD study, concerning uncertainties in numerical urban storm water drainage models. Initially an uncertainty localization and assessment of model inputs and parameters as well as uncertainties caused by different model...

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Bayesian Statistics and Uncertainty Quantification for Safety Boundary Analysis in Complex Systems

Science.gov (United States)

He, Yuning; Davies, Misty Dawn

2014-01-01

The analysis of a safety-critical system often requires detailed knowledge of safe regions and their highdimensional non-linear boundaries. We present a statistical approach to iteratively detect and characterize the boundaries, which are provided as parameterized shape candidates. Using methods from uncertainty quantification and active learning, we incrementally construct a statistical model from only few simulation runs and obtain statistically sound estimates of the shape parameters for safety boundaries.

Quantifying uncertainties in the estimation of safety parameters by using bootstrapped artificial neural networks

Energy Technology Data Exchange (ETDEWEB)

Secchi, Piercesare [MOX, Department of Mathematics, Polytechnic of Milan (Italy); Zio, Enrico [Department of Energy, Polytechnic of Milan, Via Ponzio 34/3, 20133 Milano (Italy)], E-mail: enrico.zio@polimi.it; Di Maio, Francesco [Department of Energy, Polytechnic of Milan, Via Ponzio 34/3, 20133 Milano (Italy)

2008-12-15

For licensing purposes, safety cases of Nuclear Power Plants (NPPs) must be presented at the Regulatory Authority with the necessary confidence on the models used to describe the plant safety behavior. In principle, this requires the repetition of a large number of model runs to account for the uncertainties inherent in the model description of the true plant behavior. The present paper propounds the use of bootstrapped Artificial Neural Networks (ANNs) for performing the numerous model output calculations needed for estimating safety margins with appropriate confidence intervals. Account is given both to the uncertainties inherent in the plant model and to those introduced by the ANN regression models used for performing the repeated safety parameter evaluations. The proposed framework of analysis is first illustrated with reference to a simple analytical model and then to the estimation of the safety margin on the maximum fuel cladding temperature reached during a complete group distribution header blockage scenario in a RBMK-1500 nuclear reactor. The results are compared with those obtained by a traditional parametric approach.

Coping with uncertainty in environmental impact assessments: Open techniques

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Ibsen C., E-mail: c.cardenas@utwente.nl [IceBridge Research Institutea, Universiteit Twente, P.O. Box 217, 7500 AE Enschede (Netherlands); Halman, Johannes I.M., E-mail: J.I.M.Halman@utwente.nl [Universiteit Twente, P.O. Box 217, 7500 AE Enschede (Netherlands)

2016-09-15

Uncertainty is virtually unavoidable in environmental impact assessments (EIAs). From the literature related to treating and managing uncertainty, we have identified specific techniques for coping with uncertainty in EIAs. Here, we have focused on basic steps in the decision-making process that take place within an EIA setting. More specifically, we have identified uncertainties involved in each decision-making step and discussed the extent to which these can be treated and managed in the context of an activity or project that may have environmental impacts. To further demonstrate the relevance of the techniques identified, we have examined the extent to which the EIA guidelines currently used in Colombia consider and provide guidance on managing the uncertainty involved in these assessments. Some points that should be considered in order to provide greater robustness in impact assessments in Colombia have been identified. These include the management of stakeholder values, the systematic generation of project options, and their associated impacts as well as the associated management actions, and the evaluation of uncertainties and assumptions. We believe that the relevant and specific techniques reported here can be a reference for future evaluations of other EIA guidelines in different countries. - Highlights: • uncertainty is unavoidable in environmental impact assessments, EIAs; • we have identified some open techniques to EIAs for treating and managing uncertainty in these assessments; • points for improvement that should be considered in order to provide greater robustness in EIAs in Colombia have been identified; • the paper provides substantiated a reference for possible examinations of EIAs guidelines in other countries.

Coping with uncertainty in environmental impact assessments: Open techniques

International Nuclear Information System (INIS)

Cardenas, Ibsen C.; Halman, Johannes I.M.

2016-01-01

Uncertainty is virtually unavoidable in environmental impact assessments (EIAs). From the literature related to treating and managing uncertainty, we have identified specific techniques for coping with uncertainty in EIAs. Here, we have focused on basic steps in the decision-making process that take place within an EIA setting. More specifically, we have identified uncertainties involved in each decision-making step and discussed the extent to which these can be treated and managed in the context of an activity or project that may have environmental impacts. To further demonstrate the relevance of the techniques identified, we have examined the extent to which the EIA guidelines currently used in Colombia consider and provide guidance on managing the uncertainty involved in these assessments. Some points that should be considered in order to provide greater robustness in impact assessments in Colombia have been identified. These include the management of stakeholder values, the systematic generation of project options, and their associated impacts as well as the associated management actions, and the evaluation of uncertainties and assumptions. We believe that the relevant and specific techniques reported here can be a reference for future evaluations of other EIA guidelines in different countries. - Highlights: • uncertainty is unavoidable in environmental impact assessments, EIAs; • we have identified some open techniques to EIAs for treating and managing uncertainty in these assessments; • points for improvement that should be considered in order to provide greater robustness in EIAs in Colombia have been identified; • the paper provides substantiated a reference for possible examinations of EIAs guidelines in other countries.

Risk assessment under deep uncertainty: A methodological comparison

International Nuclear Information System (INIS)

Shortridge, Julie; Aven, Terje; Guikema, Seth

2017-01-01

Probabilistic Risk Assessment (PRA) has proven to be an invaluable tool for evaluating risks in complex engineered systems. However, there is increasing concern that PRA may not be adequate in situations with little underlying knowledge to support probabilistic representation of uncertainties. As analysts and policy makers turn their attention to deeply uncertain hazards such as climate change, a number of alternatives to traditional PRA have been proposed. This paper systematically compares three diverse approaches for risk analysis under deep uncertainty (qualitative uncertainty factors, probability bounds, and robust decision making) in terms of their representation of uncertain quantities, analytical output, and implications for risk management. A simple example problem is used to highlight differences in the way that each method relates to the traditional risk assessment process and fundamental issues associated with risk assessment and description. We find that the implications for decision making are not necessarily consistent between approaches, and that differences in the representation of uncertain quantities and analytical output suggest contexts in which each method may be most appropriate. Finally, each methodology demonstrates how risk assessment can inform decision making in deeply uncertain contexts, informing more effective responses to risk problems characterized by deep uncertainty. - Highlights: • We compare three diverse approaches to risk assessment under deep uncertainty. • A simple example problem highlights differences in analytical process and results. • Results demonstrate how methodological choices can impact risk assessment results.

Interim process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Sellin, Patrick

2004-08-01

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

Interim process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrick (ed.)

2004-08-01

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

Uncertainties in life cycle assessment of waste management systems

DEFF Research Database (Denmark)

Clavreul, Julie; Christensen, Thomas Højlund

2011-01-01

Life cycle assessment has been used to assess environmental performances of waste management systems in many studies. The uncertainties inherent to its results are often pointed out but not always quantified, which should be the case to ensure a good decisionmaking process. This paper proposes...... a method to assess all parameter uncertainties and quantify the overall uncertainty of the assessment. The method is exemplified in a case study, where the goal is to determine if anaerobic digestion of organic waste is more beneficial than incineration in Denmark, considering only the impact on global...... warming. The sensitivity analysis pointed out ten parameters particularly highly influencing the result of the study. In the uncertainty analysis, the distributions of these ten parameters were used in a Monte Carlo analysis, which concluded that incineration appeared more favourable than anaerobic...

Corrosion calculations report for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

This report is a compilation of the quantitative assessments of corrosion of the copper canisters in a KBS-3 repository. The calculations are part of the safety assessment SR-Site that is the long-term safety assessment to support the license application for building a final repository for spent nuclear fuel at Forsmark, Sweden. The safety assessment methodology gives the frame for the structured and documented approach to assess all conceivable corrosion processes. The quantitative assessments are done in different ways depending on the nature of the process and on the implications for the long-term safety. The starting point for the handling of the corrosion processes is the description of all known corrosion processes for copper with the current knowledge base and applied to the specific system and geology. Already at this stage some processes are excluded for further analysis, for example if the repository environment is not a sufficient prerequisite for the process to occur. The next step is to identify processes where the extent of corrosion could be bounded, e.g. by a mass balance approach. For processes where a mass balance is not limiting, the mass transport of corrodants (or corrosion products) is taken into account. A simple approach would be just to calculate the diffusive transport of corrodants through the bentonite, but generally the transport resistance for the interface between groundwater in a rock fracture intersecting the deposition hole and the bentonite buffer is more important. In SR-Site, the concept of equivalent flowrate, Q eq , is used. This assessment is done integrated with the evaluation of the geochemical and hydrogeological evolution of the repository. For most of the corrosion processes analysed, the corrosion depth is much smaller than the copper shell thickness, even for the assessment time of 10 6 years. Several processes give corrosion depths less than 100 μm, but no process give corrosion depths larger than a few millimetres

Corrosion calculations report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This report is a compilation of the quantitative assessments of corrosion of the copper canisters in a KBS-3 repository. The calculations are part of the safety assessment SR-Site that is the long-term safety assessment to support the license application for building a final repository for spent nuclear fuel at Forsmark, Sweden. The safety assessment methodology gives the frame for the structured and documented approach to assess all conceivable corrosion processes. The quantitative assessments are done in different ways depending on the nature of the process and on the implications for the long-term safety. The starting point for the handling of the corrosion processes is the description of all known corrosion processes for copper with the current knowledge base and applied to the specific system and geology. Already at this stage some processes are excluded for further analysis, for example if the repository environment is not a sufficient prerequisite for the process to occur. The next step is to identify processes where the extent of corrosion could be bounded, e.g. by a mass balance approach. For processes where a mass balance is not limiting, the mass transport of corrodants (or corrosion products) is taken into account. A simple approach would be just to calculate the diffusive transport of corrodants through the bentonite, but generally the transport resistance for the interface between groundwater in a rock fracture intersecting the deposition hole and the bentonite buffer is more important. In SR-Site, the concept of equivalent flowrate, Q{sub eq}, is used. This assessment is done integrated with the evaluation of the geochemical and hydrogeological evolution of the repository. For most of the corrosion processes analysed, the corrosion depth is much smaller than the copper shell thickness, even for the assessment time of 106 years. Several processes give corrosion depths less than 100 mum, but no process give corrosion depths larger than a few

Uncertainty management in radioactive waste repository site assessment

International Nuclear Information System (INIS)

Baldwin, J.f.; Martin, T.P.; Tocatlidou

1994-01-01

The problem of performance assessment of a site to serve as a repository for the final disposal of radioactive waste involves different types of uncertainties. Their main sources include the large temporal and spatial considerations over which safety of the system has to be ensured, our inability to completely understand and describe a very complex structure such as the repository system, lack of precision in the measured information etc. These issues underlie most of the problems faced when rigid probabilistic approaches are used. Nevertheless a framework is needed, that would allow for an optimal aggregation of the available knowledge and an efficient management of the various types of uncertainty involved. In this work a knowledge-based modelling of the repository selection process is proposed that through a consequence analysis, evaluates the potential impact that hypothetical scenarios will have on a candidate site. The model is organised around a hierarchical structure, relating the scenarios with the possible events and processes that characterise them, and the site parameters. The scheme provides for both crisp and fuzzy parameter values and uses fuzzy semantic unification and evidential support logic reference mechanisms. It is implemented using the artificial intelligence language FRIL and the interaction with the user is performed through a windows interface

An Assessment of SKB's Performance Assessment Calculations in the Interim Main Report for the Safety Assessment SR-Can

International Nuclear Information System (INIS)

Maul, Philip; Robinson, Peter

2005-03-01

properties of the buffer and its longer-term performance. 3. The underlying methods for considering radionuclide transport are little changed from SR 97, although useful improvements have been made in some areas. The approach taken means that additional calculations are needed to address issues related to the evolution of the system with time. Whether the overall methodology will enable a comprehensive assessment to be undertaken in practice can only be judged when the full SR-Can assessment is available. 4. The documentation of the models used in PA calculations often relies on references going back over a period of twenty years updated by model validity documents for each model. The production of a single up-to-date supporting document giving full details of the models used would greatly assist the transparency of the safety case presentation. 5. The consideration of conceptual uncertainties in the supporting Process Report is restricted to the buffer. This restriction greatly limits the usefulness of the Process Report in providing information on the overall methodology. For example, it is not clear whether the approach taken for the buffer will be satisfactory for addressing conceptual model uncertainties in the geosphere. 6. SKB have not presented any deterministic PA calculations. Without these it is often difficult to understand fully the probabilistic calculations that are presented, although independent AMBER calculations have been able to reproduce the key features of these calculations. It is suggested that deterministic calculations should be part of SR-Can safety assessment. 7. It has been possible to reproduce the key features of the interim SR-Can probabilistic calculations with AMBER, although there remain uncertainties deriving from the way that SKB have modelled the U-238 decay chain in different parts of the system. 8. A full reproduction of the interim SR-Can calculations was not possible because only summary data from hydrogeological calculations are

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Uncertainties in risk assessment at USDOE facilities

Energy Technology Data Exchange (ETDEWEB)

Hamilton, L.D.; Holtzman, S.; Meinhold, A.F.; Morris, S.C.; Rowe, M.D.

1994-01-01

The United States Department of Energy (USDOE) has embarked on an ambitious program to remediate environmental contamination at its facilities. Decisions concerning cleanup goals, choices among cleanup technologies, and funding prioritization should be largely risk-based. Risk assessments will be used more extensively by the USDOE in the future. USDOE needs to develop and refine risk assessment methods and fund research to reduce major sources of uncertainty in risk assessments at USDOE facilities. The terms{open_quote} risk assessment{close_quote} and{open_quote} risk management{close_quote} are frequently confused. The National Research Council (1983) and the United States Environmental Protection Agency (USEPA, 1991a) described risk assessment as a scientific process that contributes to risk management. Risk assessment is the process of collecting, analyzing and integrating data and information to identify hazards, assess exposures and dose responses, and characterize risks. Risk characterization must include a clear presentation of {open_quotes}... the most significant data and uncertainties...{close_quotes} in an assessment. Significant data and uncertainties are {open_quotes}...those that define and explain the main risk conclusions{close_quotes}. Risk management integrates risk assessment information with other considerations, such as risk perceptions, socioeconomic and political factors, and statutes, to make and justify decisions. Risk assessments, as scientific processes, should be made independently of the other aspects of risk management (USEPA, 1991a), but current methods for assessing health risks are based on conservative regulatory principles, causing unnecessary public concern and misallocation of funds for remediation.

Uncertainties in risk assessment at USDOE facilities

International Nuclear Information System (INIS)

Hamilton, L.D.; Holtzman, S.; Meinhold, A.F.; Morris, S.C.; Rowe, M.D.

1994-01-01

The United States Department of Energy (USDOE) has embarked on an ambitious program to remediate environmental contamination at its facilities. Decisions concerning cleanup goals, choices among cleanup technologies, and funding prioritization should be largely risk-based. Risk assessments will be used more extensively by the USDOE in the future. USDOE needs to develop and refine risk assessment methods and fund research to reduce major sources of uncertainty in risk assessments at USDOE facilities. The terms open-quote risk assessment close-quote and open-quote risk management close-quote are frequently confused. The National Research Council (1983) and the United States Environmental Protection Agency (USEPA, 1991a) described risk assessment as a scientific process that contributes to risk management. Risk assessment is the process of collecting, analyzing and integrating data and information to identify hazards, assess exposures and dose responses, and characterize risks. Risk characterization must include a clear presentation of open-quotes... the most significant data and uncertainties...close quotes in an assessment. Significant data and uncertainties are open-quotes...those that define and explain the main risk conclusionsclose quotes. Risk management integrates risk assessment information with other considerations, such as risk perceptions, socioeconomic and political factors, and statutes, to make and justify decisions. Risk assessments, as scientific processes, should be made independently of the other aspects of risk management (USEPA, 1991a), but current methods for assessing health risks are based on conservative regulatory principles, causing unnecessary public concern and misallocation of funds for remediation

Decay heat uncertainty quantification of MYRRHA

Directory of Open Access Journals (Sweden)

Fiorito Luca

2017-01-01

Full Text Available MYRRHA is a lead-bismuth cooled MOX-fueled accelerator driven system (ADS currently in the design phase at SCK·CEN in Belgium. The correct evaluation of the decay heat and of its uncertainty level is very important for the safety demonstration of the reactor. In the first part of this work we assessed the decay heat released by the MYRRHA core using the ALEPH-2 burnup code. The second part of the study focused on the nuclear data uncertainty and covariance propagation to the MYRRHA decay heat. Radioactive decay data, independent fission yield and cross section uncertainties/covariances were propagated using two nuclear data sampling codes, namely NUDUNA and SANDY. According to the results, 238U cross sections and fission yield data are the largest contributors to the MYRRHA decay heat uncertainty. The calculated uncertainty values are deemed acceptable from the safety point of view as they are well within the available regulatory limits.

Assessment of volcanic hazards, vulnerability, risk and uncertainty (Invited)

Science.gov (United States)

Sparks, R. S.

2009-12-01

many sources of uncertainty in forecasting the areas that volcanic activity will effect and the severity of the effects. Uncertainties arise from: natural variability, inadequate data, biased data, incomplete data, lack of understanding of the processes, limitations to predictive models, ambiguity, and unknown unknowns. The description of volcanic hazards is thus necessarily probabilistic and requires assessment of the attendant uncertainties. Several issues arise from the probabilistic nature of volcanic hazards and the intrinsic uncertainties. Although zonation maps require well-defined boundaries for administrative pragmatism, such boundaries cannot divide areas that are completely safe from those that are unsafe. Levels of danger or safety need to be defined to decide on and justify boundaries through the concepts of vulnerability and risk. More data, better observations, improved models may reduce uncertainties, but can increase uncertainties and may lead to re-appraisal of zone boundaries. Probabilities inferred by statistical techniques are hard to communicate. Expert elicitation is an emerging methodology for risk assessment and uncertainty evaluation. The method has been applied at one major volcanic crisis (Soufrière Hills Volcano, Montserrat), and is being applied in planning for volcanic crises at Vesuvius.

Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Eom, H.S.; Sung, T.Y.; Jeong, H.S.; Park, J.H.; Kang, H.G.; Lee, K

2001-03-01

As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software.

Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

International Nuclear Information System (INIS)

Eom, H. S.; Sung, T. Y.; Jeong, H. S.; Park, J. H.; Kang, H. G.; Lee, K.

2001-03-01

As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software

Establishment of safety goal and its quantification based on risk assessment

International Nuclear Information System (INIS)

Miyano, Hiroshi; Muramatsu, Ken

2017-01-01

We must clarify the safety objectives sought by society in securing the safety of nuclear reactors and nuclear power plants. For that purpose, it is useful to utilize risk assessment. Quantitative methods including probabilistic risk assessment (PRA) are superior in terms of scientific rationality and quantitative performance compared with conventional deterministic methods, and able to indicate an objective numerical value of safety level. Consequently, quantitative methods can enhance the transparency, consistency, compliance, predictability, and explanatory power of regulatory decisions toward business operators and citizens. Business operators can explain the validity of their own safety assurance activities to regulators and citizens. The goal to be secured becomes clear by incorporating the safety goal into the specific performance goal required for the nuclear power plant from the viewpoint of deep safeguard, and it becomes easy to evaluate the effectiveness of the safety measures. It helps us greatly in judging and selecting the appropriateness of safety measures. It should be noted: the fact that the result of implementing the PRA satisfies the safety goal is not a sufficient condition in the sense of guaranteeing complete safety but a necessary condition. The nuclear power field is a region with large uncertainty, and research/efforts for accuracy improvement and evaluation validity will be required continuously. (A.O.)

CEC/USDOE workshop on uncertainty analysis

International Nuclear Information System (INIS)

Elderkin, C.E.; Kelly, G.N.

1990-07-01

Any measured or assessed quantity contains uncertainty. The quantitative estimation of such uncertainty is becoming increasingly important, especially in assuring that safety requirements are met in design, regulation, and operation of nuclear installations. The CEC/USDOE Workshop on Uncertainty Analysis, held in Santa Fe, New Mexico, on November 13 through 16, 1989, was organized jointly by the Commission of European Communities (CEC's) Radiation Protection Research program, dealing with uncertainties throughout the field of consequence assessment, and DOE's Atmospheric Studies in Complex Terrain (ASCOT) program, concerned with the particular uncertainties in time and space variant transport and dispersion. The workshop brought together US and European scientists who have been developing or applying uncertainty analysis methodologies, conducted in a variety of contexts, often with incomplete knowledge of the work of others in this area. Thus, it was timely to exchange views and experience, identify limitations of approaches to uncertainty and possible improvements, and enhance the interface between developers and users of uncertainty analysis methods. Furthermore, the workshop considered the extent to which consistent, rigorous methods could be used in various applications within consequence assessment. 3 refs

Technical Standards on the Safety Assessment of a HLW Repository in Other Countries

International Nuclear Information System (INIS)

Lee, Sung Ho; Hwang, Yong Soo

2009-01-01

The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

Directory of Open Access Journals (Sweden)

Herrero J.J.

2017-01-01

Full Text Available In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

Uncertainty of a hydrological climate change impact assessment - Is it really all about climate uncertainty?

Science.gov (United States)

Honti, Mark; Reichert, Peter; Scheidegger, Andreas; Stamm, Christian

2013-04-01

Climate change impact assessments have become more and more popular in hydrology since the middle 1980's with another boost after the publication of the IPCC AR4 report. During hundreds of impact studies a quasi-standard methodology emerged, which is mainly shaped by the growing public demand for predicting how water resources management or flood protection should change in the close future. The ``standard'' workflow considers future climate under a specific IPCC emission scenario simulated by global circulation models (GCMs), possibly downscaled by a regional climate model (RCM) and/or a stochastic weather generator. The output from the climate models is typically corrected for bias before feeding it into a calibrated hydrological model, which is run on the past and future meteorological data to analyse the impacts of climate change on the hydrological indicators of interest. The impact predictions are as uncertain as any forecast that tries to describe the behaviour of an extremely complex system decades into the future. Future climate predictions are uncertain due to the scenario uncertainty and the GCM model uncertainty that is obvious on finer resolution than continental scale. Like in any hierarchical model system, uncertainty propagates through the descendant components. Downscaling increases uncertainty with the deficiencies of RCMs and/or weather generators. Bias correction adds a strong deterministic shift to the input data. Finally the predictive uncertainty of the hydrological model ends the cascade that leads to the total uncertainty of the hydrological impact assessment. There is an emerging consensus between many studies on the relative importance of the different uncertainty sources. The prevailing perception is that GCM uncertainty dominates hydrological impact studies. There are only few studies, which found that the predictive uncertainty of hydrological models can be in the same range or even larger than climatic uncertainty. We carried out a

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)

The costs of uncertainty: regulating health and safety in the Canadian uranium industry

International Nuclear Information System (INIS)

Robinson, I.

1982-04-01

Federalism, and particularly federal/provincial jurisdictional relationships, have led to considerable uncertainty in the regulation of occupational health and safety and of environmental protection in the Canadian uranium mining industry. The two principal uranium producing provinces in Canada are Saskatchewan and Ontario. Since 1978, in an attempt to avoid constitutional issues, both these provinces and the federal government as well have proceeded unilaterally with health and safety reforms for the industry. In Saskatchewan this has resulted in areas of overlapping jurisdiction, which have led to uncertainty over the legal enforceability of the provincial regulations. In Ontario, the province has left significant gaps in the protection of both workers and the environment. Little progress can be expected in eliminating these gaps and overlaps until the current administrative and jurisdictional arrangements are understood

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

International Nuclear Information System (INIS)

Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

2013-01-01

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

Spatial variability and parametric uncertainty in performance assessment models

International Nuclear Information System (INIS)

Pensado, Osvaldo; Mancillas, James; Painter, Scott; Tomishima, Yasuo

2011-01-01

The problem of defining an appropriate treatment of distribution functions (which could represent spatial variability or parametric uncertainty) is examined based on a generic performance assessment model for a high-level waste repository. The generic model incorporated source term models available in GoldSim ® , the TDRW code for contaminant transport in sparse fracture networks with a complex fracture-matrix interaction process, and a biosphere dose model known as BDOSE TM . Using the GoldSim framework, several Monte Carlo sampling approaches and transport conceptualizations were evaluated to explore the effect of various treatments of spatial variability and parametric uncertainty on dose estimates. Results from a model employing a representative source and ensemble-averaged pathway properties were compared to results from a model allowing for stochastic variation of transport properties along streamline segments (i.e., explicit representation of spatial variability within a Monte Carlo realization). We concluded that the sampling approach and the definition of an ensemble representative do influence consequence estimates. In the examples analyzed in this paper, approaches considering limited variability of a transport resistance parameter along a streamline increased the frequency of fast pathways resulting in relatively high dose estimates, while those allowing for broad variability along streamlines increased the frequency of 'bottlenecks' reducing dose estimates. On this basis, simplified approaches with limited consideration of variability may suffice for intended uses of the performance assessment model, such as evaluation of site safety. (author)

Construction of knowledge base for geological disposal technologies of high-level radioactive waste Report in 2005. Separate volume 3: Development of safety assessment methods

International Nuclear Information System (INIS)

2005-09-01

The results of development of safety assessment methods by JNC after the second report are reported. JNC-Thermodynamic and JNC-Sorption Database of nuclides was prepared and used. The mass transfer process model in rock, the water quality model of underwater and pore water and approach to modeling radionuclide transport in biosphere were improved. The phenomenological nuclide transport model and the effect assessment model of colloid, natural organic compounds and microorganism were developed. On scenario of safety assessment method, the behaviors in the disposal system were expressed by FEP (Features, Events, and Processes). The effects of data uncertainty and model uncertainty were improved by the assessment technologies and the sensitivity analysis technology. JGIS (JNC Geological Disposal Information Integration System) was developed. The main performance of JGIS was shown. It consists of six chapters; the first chapter is introduction, the second chapter the nuclides transport database, the third the safety assessment model, the forth improvement of safety assessment methods, the fifth application of safety assessment methods and the sixth results and summary. (S.Y.)

Coping with uncertainty in environmental impact assessments: Open techniques

NARCIS (Netherlands)

Chivatá Cárdenas, Ibsen; Halman, Johannes I.M.

2016-01-01

Uncertainty is virtually unavoidable in environmental impact assessments (EIAs). From the literature related to treating and managing uncertainty, we have identified specific techniques for coping with uncertainty in EIAs. Here, we have focused on basic steps in the decision-making process that take

Assessing uncertainty and risk in exploited marine populations

International Nuclear Information System (INIS)

Fogarty, M.J.; Mayo, R.K.; O'Brien, L.; Serchuk, F.M.; Rosenberg, A.A.

1996-01-01

The assessment and management of exploited fish and invertebrate populations is subject to several types of uncertainty. This uncertainty translates into risk to the population in the development and implementation of fishery management advice. Here, we define risk as the probability that exploitation rates will exceed a threshold level where long term sustainability of the stock is threatened. We distinguish among several sources of error or uncertainty due to (a) stochasticity in demographic rates and processes, particularly in survival rates during the early fife stages; (b) measurement error resulting from sampling variation in the determination of population parameters or in model estimation; and (c) the lack of complete information on population and ecosystem dynamics. The first represents a form of aleatory uncertainty while the latter two factors represent forms of epistemic uncertainty. To illustrate these points, we evaluate the recent status of the Georges Bank cod stock in a risk assessment framework. Short term stochastic projections are made accounting for uncertainty in population size and for random variability in the number of young surviving to enter the fishery. We show that recent declines in this cod stock can be attributed to exploitation rates that have substantially exceeded sustainable levels

Risk-based approach to long-term safety assessment for near surface disposal of radioactive waste in Korea

International Nuclear Information System (INIS)

Jeong, C.W.; Kim, K.I.; Lee, J.I.

2000-01-01

This paper presents the Korean regulatory approach to safety assessment consistent with probabilistic, risk-based long-term safety requirements for near surface disposal facilities. The approach is based on: (1) From the standpoint of risk limitation, normal processes and probabilistic disruptive events should be integrated in a similar manner in terms of potential exposures; and (2) The uncertainties inherent in the safety assessment should be reduced using appropriate exposure scenarios. In addition, this paper emphasizes the necessity of international guidance for quantifying potential exposures and the corresponding risks from radioactive waste disposal. (author)

Risk assessment through drinking water pathway via uncertainty modeling of contaminant transport using soft computing

International Nuclear Information System (INIS)

Datta, D.; Ranade, A.K.; Pandey, M.; Sathyabama, N.; Kumar, Brij

2012-01-01

The basic objective of an environmental impact assessment (EIA) is to build guidelines to reduce the associated risk or mitigate the consequences of the reactor accident at its source to prevent deterministic health effects, to reduce the risk of stochastic health effects (eg. cancer and severe hereditary effects) as much as reasonable achievable by implementing protective actions in accordance with IAEA guidance (IAEA Safety Series No. 115, 1996). The measure of exposure being the basic tool to take any appropriate decisions related to risk reduction, EIA is traditionally expressed in terms of radiation exposure to the member of the public. However, models used to estimate the exposure received by the member of the public are governed by parameters some of which are deterministic with relative uncertainty and some of which are stochastic as well as imprecise (insufficient knowledge). In an admixture environment of this type, it is essential to assess the uncertainty of a model to estimate the bounds of the exposure to the public to invoke a decision during an event of nuclear or radiological emergency. With a view to this soft computing technique such as evidence theory based assessment of model parameters is addressed to compute the risk or exposure to the member of the public. The possible pathway of exposure to the member of the public in the aquatic food stream is the drinking of water. Accordingly, this paper presents the uncertainty analysis of exposure via uncertainty analysis of the contaminated water. Evidence theory finally addresses the uncertainty in terms of lower bound as belief measure and upper bound of exposure as plausibility measure. In this work EIA is presented using evidence theory. Data fusion technique is used to aggregate the knowledge on the uncertain information. Uncertainty of concentration and exposure is expressed as an interval of belief, plausibility

Vector network analyzer (VNA) measurements and uncertainty assessment

CERN Document Server

Shoaib, Nosherwan

2017-01-01

This book describes vector network analyzer measurements and uncertainty assessments, particularly in waveguide test-set environments, in order to establish their compatibility to the International System of Units (SI) for accurate and reliable characterization of communication networks. It proposes a fully analytical approach to measurement uncertainty evaluation, while also highlighting the interaction and the linear propagation of different uncertainty sources to compute the final uncertainties associated with the measurements. The book subsequently discusses the dimensional characterization of waveguide standards and the quality of the vector network analyzer (VNA) calibration techniques. The book concludes with an in-depth description of the novel verification artefacts used to assess the performance of the VNAs. It offers a comprehensive reference guide for beginners to experts, in both academia and industry, whose work involves the field of network analysis, instrumentation and measurements.

Uncertainty Assessment: What Good Does it Do? (Invited)

Science.gov (United States)

Oreskes, N.; Lewandowsky, S.

2013-12-01

the public debate or advance public policy. We argue that attempts to address public doubts by improving uncertainty assessment are bound to fail, insofar as the motives for doubt-mongering are independent of scientific uncertainty, and therefore remain unaffected even as those uncertainties are diminished. We illustrate this claim by consideration of the evolution of the debate over the past ten years over the relationship between hurricanes and anthropogenic climate change. We suggest that scientists should pursue uncertainty assessment if such assessment improves scientific understanding, but not as a means to reduce public doubts or advance public policy in relation to anthropogenic climate change.

Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

Energy Technology Data Exchange (ETDEWEB)

Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

1997-06-01

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

International Nuclear Information System (INIS)

Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

1997-06-01

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses

Scenario Analysis for the Safety Assessment of Nuclear Waste Repositories: A Critical Review.

Science.gov (United States)

Tosoni, Edoardo; Salo, Ahti; Zio, Enrico

2018-04-01

A major challenge in scenario analysis for the safety assessment of nuclear waste repositories pertains to the comprehensiveness of the set of scenarios selected for assessing the safety of the repository. Motivated by this challenge, we discuss the aspects of scenario analysis relevant to comprehensiveness. Specifically, we note that (1) it is necessary to make it clear why scenarios usually focus on a restricted set of features, events, and processes; (2) there is not yet consensus on the interpretation of comprehensiveness for guiding the generation of scenarios; and (3) there is a need for sound approaches to the treatment of epistemic uncertainties. © 2017 Society for Risk Analysis.

Bedrock K{sub d} data and uncertainty assessment for application in SR-Site geosphere transport calculations

Energy Technology Data Exchange (ETDEWEB)

Crawford, James (Kemakta Konsult AB, Stockholm (Sweden))

2010-12-15

The safety assessment SR-Site is undertaken to assess the safety of a potential geologic repository for spent nuclear fuel at the Forsmark and Laxemar sites. The present report is one of several reports that form the data input to SR-Site and contains a compilation of recommended K{sub d} data (i.e. linear partitioning coefficients) for safety assessment modelling of geosphere radionuclide transport. The data are derived for rock types and groundwater compositions distinctive of the site investigation areas at Forsmark and Laxemar. Data have been derived for all elements and redox states considered of importance for far-field dose estimates as described in /SKB 2010d/. The K{sub d} data are given in the form of lognormal distributions characterised by a mean (mu) and standard deviation (sigma). Upper and lower limits for the uncertainty range of the recommended data are defined by the 2.5% and 97.5% percentiles of the empirical data sets. The best estimate K{sub d} value for use in deterministic calculations is given as the median of the K{sub d} distribution

Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis

Energy Technology Data Exchange (ETDEWEB)

Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

2017-11-15

The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

One Approach to the Fire PSA Uncertainty Analysis

International Nuclear Information System (INIS)

Simic, Z.; Mikulicic, V.; Vukovic, I.

2002-01-01

Experienced practical events and findings from the number of fire probabilistic safety assessment (PSA) studies show that fire has high relative importance for nuclear power plant safety. Fire PSA is a very challenging phenomenon and a number of issues are still in the area of research and development. This has a major impact on the conservatism of fire PSA findings. One way to reduce the level of conservatism is to conduct uncertainty analysis. At the top-level, uncertainty of the fire PSA can be separated in to three segments. The first segment is related to fire initiating events frequencies. The second uncertainty segment is connected to the uncertainty of fire damage. Finally, there is uncertainty related to the PSA model, which propagates this fire-initiated damage to the core damage or other analyzed risk. This paper discusses all three segments of uncertainty. Some recent experience with fire PSA study uncertainty analysis, usage of fire analysis code COMPBRN IIIe, and uncertainty evaluation importance to the final result is presented.(author)

Avoiding climate change uncertainties in Strategic Environmental Assessment

DEFF Research Database (Denmark)

Larsen, Sanne Vammen; Kørnøv, Lone; Driscoll, Patrick Arthur

2013-01-01

This article is concerned with how Strategic Environmental Assessment (SEA) practice handles climate change uncertainties within the Danish planning system. First, a hypothetical model is set up for how uncertainty is handled and not handled in decision-making. The model incorporates the strategies...

Sensitivity and uncertainty analyses for performance assessment modeling

International Nuclear Information System (INIS)

Doctor, P.G.

1988-08-01

Sensitivity and uncertainty analyses methods for computer models are being applied in performance assessment modeling in the geologic high level radioactive waste repository program. The models used in performance assessment tend to be complex physical/chemical models with large numbers of input variables. There are two basic approaches to sensitivity and uncertainty analyses: deterministic and statistical. The deterministic approach to sensitivity analysis involves numerical calculation or employs the adjoint form of a partial differential equation to compute partial derivatives; the uncertainty analysis is based on Taylor series expansions of the input variables propagated through the model to compute means and variances of the output variable. The statistical approach to sensitivity analysis involves a response surface approximation to the model with the sensitivity coefficients calculated from the response surface parameters; the uncertainty analysis is based on simulation. The methods each have strengths and weaknesses. 44 refs

Communicating uncertainties in assessments of future sea level rise

Science.gov (United States)

Wikman-Svahn, P.

2013-12-01

How uncertainty should be managed and communicated in policy-relevant scientific assessments is directly connected to the role of science and the responsibility of scientists. These fundamentally philosophical issues influence how scientific assessments are made and how scientific findings are communicated to policymakers. It is therefore of high importance to discuss implicit assumptions and value judgments that are made in policy-relevant scientific assessments. The present paper examines these issues for the case of scientific assessments of future sea level rise. The magnitude of future sea level rise is very uncertain, mainly due to poor scientific understanding of all physical mechanisms affecting the great ice sheets of Greenland and Antarctica, which together hold enough land-based ice to raise sea levels more than 60 meters if completely melted. There has been much confusion from policymakers on how different assessments of future sea levels should be interpreted. Much of this confusion is probably due to how uncertainties are characterized and communicated in these assessments. The present paper draws on the recent philosophical debate on the so-called "value-free ideal of science" - the view that science should not be based on social and ethical values. Issues related to how uncertainty is handled in scientific assessments are central to this debate. This literature has much focused on how uncertainty in data, parameters or models implies that choices have to be made, which can have social consequences. However, less emphasis has been on how uncertainty is characterized when communicating the findings of a study, which is the focus of the present paper. The paper argues that there is a tension between on the one hand the value-free ideal of science and on the other hand usefulness for practical applications in society. This means that even if the value-free ideal could be upheld in theory, by carefully constructing and hedging statements characterizing

The role of uncertainty analysis in dose reconstruction and risk assessment

International Nuclear Information System (INIS)

Hoffman, F.O.; Simon, S.L.; Thiessen. K.M.

1996-01-01

Dose reconstruction and risk assessment rely heavily on the use of mathematical models to extrapolate information beyond the realm of direct observation. Because models are merely approximations of real systems, their predictions are inherently uncertain. As a result, full disclosure of uncertainty in dose and risk estimates is essential to achieve scientific credibility and to build public trust. The need for formal analysis of uncertainty in model predictions was presented during the nineteenth annual meeting of the NCRP. At that time, quantitative uncertainty analysis was considered a relatively new and difficult subject practiced by only a few investigators. Today, uncertainty analysis has become synonymous with the assessment process itself. When an uncertainty analysis is used iteratively within the assessment process, it can guide experimental research to refine dose and risk estimates, deferring potentially high cost or high consequence decisions until uncertainty is either acceptable or irreducible. Uncertainty analysis is now mandated for all ongoing dose reconstruction projects within the United States, a fact that distinguishes dose reconstruction from other types of exposure and risk assessments. 64 refs., 6 figs., 1 tab

Concrete structures. Contribution to the safety assessment of existing structures

Directory of Open Access Journals (Sweden)

D. COUTO

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

Uncertainty and sensitivity analysis in nuclear accident consequence assessment

International Nuclear Information System (INIS)

Karlberg, Olof.

1989-01-01

This report contains the results of a four year project in research contracts with the Nordic Cooperation in Nuclear Safety and the National Institute for Radiation Protection. An uncertainty/sensitivity analysis methodology consisting of Latin Hypercube sampling and regression analysis was applied to an accident consequence model. A number of input parameters were selected and the uncertainties related to these parameter were estimated within a Nordic group of experts. Individual doses, collective dose, health effects and their related uncertainties were then calculated for three release scenarios and for a representative sample of meteorological situations. From two of the scenarios the acute phase after an accident were simulated and from one the long time consequences. The most significant parameters were identified. The outer limits of the calculated uncertainty distributions are large and will grow to several order of magnitudes for the low probability consequences. The uncertainty in the expectation values are typical a factor 2-5 (1 Sigma). The variation in the model responses due to the variation of the weather parameters is fairly equal to the parameter uncertainty induced variation. The most important parameters showed out to be different for each pathway of exposure, which could be expected. However, the overall most important parameters are the wet deposition coefficient and the shielding factors. A general discussion of the usefulness of uncertainty analysis in consequence analysis is also given. (au)

Structural safety - Is the safety margin measurable

International Nuclear Information System (INIS)

Rintamaa, R.

1992-01-01

In ensuring the structural safety of the nuclear components one must be aware of the uncertainties related to the material deorientation, loadings and other operational conditions, geometrical dimensions as well as the service environment. Furthermore, the validation of the analysis tools and procedures is of great importance in overall safety assessment of a pressure retaining component. In order to identify and quantify the concerns and risks arising from the uncertainties in the safety related issue intensive research is being carried out all over the world, in particular, on the ageing, plant life extension and management of old nuclear power plants. The presentation includes a general survey of the factors relevant to the assessment of safe and reliable operation of a nuclear component throughout its planned service life. Certain aspects are outlined based on the research work being carried out at the Technical Research Centre of Finland (VTT)(orig.)

Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

Energy Technology Data Exchange (ETDEWEB)

Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

Safety Assessment for Decommissioning

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

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

Quantification of margins and uncertainties: Alternative representations of epistemic uncertainty

International Nuclear Information System (INIS)

Helton, Jon C.; Johnson, Jay D.

2011-01-01

In 2001, the National Nuclear Security Administration of the U.S. Department of Energy in conjunction with the national security laboratories (i.e., Los Alamos National Laboratory, Lawrence Livermore National Laboratory and Sandia National Laboratories) initiated development of a process designated Quantification of Margins and Uncertainties (QMU) for the use of risk assessment methodologies in the certification of the reliability and safety of the nation's nuclear weapons stockpile. A previous presentation, 'Quantification of Margins and Uncertainties: Conceptual and Computational Basis,' describes the basic ideas that underlie QMU and illustrates these ideas with two notional examples that employ probability for the representation of aleatory and epistemic uncertainty. The current presentation introduces and illustrates the use of interval analysis, possibility theory and evidence theory as alternatives to the use of probability theory for the representation of epistemic uncertainty in QMU-type analyses. The following topics are considered: the mathematical structure of alternative representations of uncertainty, alternative representations of epistemic uncertainty in QMU analyses involving only epistemic uncertainty, and alternative representations of epistemic uncertainty in QMU analyses involving a separation of aleatory and epistemic uncertainty. Analyses involving interval analysis, possibility theory and evidence theory are illustrated with the same two notional examples used in the presentation indicated above to illustrate the use of probability to represent aleatory and epistemic uncertainty in QMU analyses.

Assessing framing of uncertainties in water management practice

NARCIS (Netherlands)

Isendahl, N.; Dewulf, A.; Brugnach, M.; Francois, G.; Möllenkamp, S.; Pahl-Wostl, C.

2009-01-01

Dealing with uncertainties in water management is an important issue and is one which will only increase in light of global changes, particularly climate change. So far, uncertainties in water management have mostly been assessed from a scientific point of view, and in quantitative terms. In this

Estimating uncertainty of data limited stock assessments

DEFF Research Database (Denmark)

Kokkalis, Alexandros; Eikeset, Anne Maria; Thygesen, Uffe Høgsbro

2017-01-01

-limited. Particular emphasis is put on providing uncertainty estimates of the data-limited assessment. We assess four cod stocks in the North-East Atlantic and compare our estimates of stock status (F/Fmsy) with the official assessments. The estimated stock status of all four cod stocks followed the established stock...

Methods for Assessing Uncertainties in Climate Change, Impacts and Responses (Invited)

Science.gov (United States)

Manning, M. R.; Swart, R.

2009-12-01

Assessing the scientific uncertainties or confidence levels for the many different aspects of climate change is particularly important because of the seriousness of potential impacts and the magnitude of economic and political responses that are needed to mitigate climate change effectively. This has made the treatment of uncertainty and confidence a key feature in the assessments carried out by the Intergovernmental Panel on Climate Change (IPCC). Because climate change is very much a cross-disciplinary area of science, adequately dealing with uncertainties requires recognition of their wide range and different perspectives on assessing and communicating those uncertainties. The structural differences that exist across disciplines are often embedded deeply in the corresponding literature that is used as the basis for an IPCC assessment. The assessment of climate change science by the IPCC has from its outset tried to report the levels of confidence and uncertainty in the degree of understanding in both the underlying multi-disciplinary science and in projections for future climate. The growing recognition of the seriousness of this led to the formation of a detailed approach for consistent treatment of uncertainties in the IPCC’s Third Assessment Report (TAR) [Moss and Schneider, 2000]. However, in completing the TAR there remained some systematic differences between the disciplines raising concerns about the level of consistency. So further consideration of a systematic approach to uncertainties was undertaken for the Fourth Assessment Report (AR4). The basis for the approach used in the AR4 was developed at an expert meeting of scientists representing many different disciplines. This led to the introduction of a broader way of addressing uncertainties in the AR4 [Manning et al., 2004] which was further refined by lengthy discussions among many IPCC Lead Authors, for over a year, resulting in a short summary of a standard approach to be followed for that

Treating Uncertainties in A Nuclear Seismic Probabilistic Risk Assessment by Means of the Distemper-Safer Theory of Evidence

International Nuclear Information System (INIS)

Lo, Chungkung; Pedroni, N.; Zio, E.

2014-01-01

The analyses carried out within the Seismic Probabilistic Risk Assessments (SPRAs) of Nuclear Power Plants (NPPs) are affected by significant aleatory and epistemic uncertainties. These uncertainties have to be represented and quantified coherently with the data, information and knowledge available, to provide reasonable assurance that related decisions can be taken robustly and with confidence. The amount of data, information and knowledge available for seismic risk assessment is typically limited, so that the analysis must strongly rely on expert judgments. In this paper, a Dempster-Shafer Theory (DST) framework for handling uncertainties in NPP SPRAs is proposed and applied to an example case study. The main contributions of this paper are two: (i) applying the complete DST framework to SPRA models, showing how to build the Dempster-Shafer structures of the uncertainty parameters based on industry generic data, and (ii) embedding Bayesian updating based on plant specific data into the framework. The results of the application to a case study show that the approach is feasible and effective in (i) describing and jointly propagating aleatory and epistemic uncertainties in SPRA models and (ii) providing 'conservative' bounds on the safety quantities of interest (i. e. Core Damage Frequency, CDF) that reflect the (limited) state of knowledge of the experts about the system of interest

Treating Uncertainties in A Nuclear Seismic Probabilistic Risk Assessment by Means of the Distemper-Safer Theory of Evidence

Energy Technology Data Exchange (ETDEWEB)

Lo, Chungkung [Chair on Systems Science and the Energetic Challenge, Paris (France); Pedroni, N.; Zio, E. [Politecnico di Milano, Milano (Italy)

2014-02-15

The analyses carried out within the Seismic Probabilistic Risk Assessments (SPRAs) of Nuclear Power Plants (NPPs) are affected by significant aleatory and epistemic uncertainties. These uncertainties have to be represented and quantified coherently with the data, information and knowledge available, to provide reasonable assurance that related decisions can be taken robustly and with confidence. The amount of data, information and knowledge available for seismic risk assessment is typically limited, so that the analysis must strongly rely on expert judgments. In this paper, a Dempster-Shafer Theory (DST) framework for handling uncertainties in NPP SPRAs is proposed and applied to an example case study. The main contributions of this paper are two: (i) applying the complete DST framework to SPRA models, showing how to build the Dempster-Shafer structures of the uncertainty parameters based on industry generic data, and (ii) embedding Bayesian updating based on plant specific data into the framework. The results of the application to a case study show that the approach is feasible and effective in (i) describing and jointly propagating aleatory and epistemic uncertainties in SPRA models and (ii) providing 'conservative' bounds on the safety quantities of interest (i. e. Core Damage Frequency, CDF) that reflect the (limited) state of knowledge of the experts about the system of interest.

Knowledge representation in safety assessment: improving transparency and traceability

Energy Technology Data Exchange (ETDEWEB)

Lemos, F.L. de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Sullivan, T. [Brookhaven National Laboratory (BNL), Upton, NY (United States); Ross, T. [University of New Mexico (UNM), Albuquerque, NM (United States); Guimaraes, L.N.F. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)

2011-07-01

Transparency and traceability are key factors for confidence building, acceptability, and quality enhancement of the safety assessment, and safety case for a radioactive waste disposal facility. In order to facilitate analysis and promote discussions, all of the information used to make decisions should be readily available to stake holders. The information should convey a good understanding of the intermediate decisions processes, allowing examination of alternatives and 'what if questions'. In an ideal situation all stake holders, including scientists and the public, should be able to follow the path of a certain parameter, from the beginning where it was defined, its assumptions and uncertainties, throughout the calculations until the final results of the safety assessment. One of the main challenges, to achieving such a transparency and traceability, is that stake holders are a very diverse audience, with very different backgrounds. This could require preparation of various versions of the same documentation, which would be impractical. While the linguistic information is of crucial importance to understanding the reasoning, it is very difficult to convey the supporting conditions, and consequent uncertainties for the selection of parameters values. Even scientists involved in the process can become confused due to the overwhelming amount of information that is used to support parameter value selection. The amount of details makes it difficult to track the decisions, which lead to the selection of a certain parameter, throughout the calculations. This paper presents a methodology to represent the linguistic information used in the safety assessment in terms of mathematical expressions by using the fuzzy sets and fuzzy logic tools. This methodology aims to help information to be readily available while keeping, as much as possible, the original meaning of the linguistic expressions and, consequently, to be available at any time as a quick reference

Knowledge representation in safety assessment: improving transparency and traceability

International Nuclear Information System (INIS)

Lemos, F.L. de; Sullivan, T.; Ross, T.; Guimaraes, L.N.F.

2011-01-01

Transparency and traceability are key factors for confidence building, acceptability, and quality enhancement of the safety assessment, and safety case for a radioactive waste disposal facility. In order to facilitate analysis and promote discussions, all of the information used to make decisions should be readily available to stake holders. The information should convey a good understanding of the intermediate decisions processes, allowing examination of alternatives and 'what if questions'. In an ideal situation all stake holders, including scientists and the public, should be able to follow the path of a certain parameter, from the beginning where it was defined, its assumptions and uncertainties, throughout the calculations until the final results of the safety assessment. One of the main challenges, to achieving such a transparency and traceability, is that stake holders are a very diverse audience, with very different backgrounds. This could require preparation of various versions of the same documentation, which would be impractical. While the linguistic information is of crucial importance to understanding the reasoning, it is very difficult to convey the supporting conditions, and consequent uncertainties for the selection of parameters values. Even scientists involved in the process can become confused due to the overwhelming amount of information that is used to support parameter value selection. The amount of details makes it difficult to track the decisions, which lead to the selection of a certain parameter, throughout the calculations. This paper presents a methodology to represent the linguistic information used in the safety assessment in terms of mathematical expressions by using the fuzzy sets and fuzzy logic tools. This methodology aims to help information to be readily available while keeping, as much as possible, the original meaning of the linguistic expressions and, consequently, to be available at any time as a quick reference. This would

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

International Nuclear Information System (INIS)

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

2008-12-01

scope of the quantitative safety assessment. These arguments include: Support from natural and anthropogenic analogues for both key process understanding and total system performance. Comparison of the methodology and results with the earlier TILA-99 and SR-Can safety assessments, as well as other international safety assessments, to ensure completeness, consistency and reasonableness of the present assessment. Use of safety indicators other than dose and activity to avoid uncertainties in future human lifestyles and also in geological processes on very long timescales. Consideration of the calculation results from a wider perspective to consider significance of their impact compared to other risks

ESFR core optimization and uncertainty studies

International Nuclear Information System (INIS)

Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

2015-01-01

In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR

Uncertainty propagation in probabilistic risk assessment: A comparative study

International Nuclear Information System (INIS)

Ahmed, S.; Metcalf, D.R.; Pegram, J.W.

1982-01-01

Three uncertainty propagation techniques, namely method of moments, discrete probability distribution (DPD), and Monte Carlo simulation, generally used in probabilistic risk assessment, are compared and conclusions drawn in terms of the accuracy of the results. For small uncertainty in the basic event unavailabilities, the three methods give similar results. For large uncertainty, the method of moments is in error, and the appropriate method is to propagate uncertainty in the discrete form either by DPD method without sampling or by Monte Carlo. (orig.)

Uncertainty representation and combination: new results with application to nuclear safety issues

International Nuclear Information System (INIS)

Destercke, S.

2008-10-01

It often happens that the value of some parameters or variables of a system are imperfectly known, either because of the variability of the modelled phenomena, or because the available information is imprecise or incomplete. Classical probability theory is usually used to treat these uncertainties. However, recent years have witnessed the appearance of arguments pointing to the conclusion that classical probabilities are inadequate to handle imprecise or incomplete information. Other frameworks have thus been proposed to address this problem: the three main are probability sets, random sets and possibility theory. There are many open questions concerning uncertainty treatment within these frameworks. More precisely, it is necessary to build bridges between these three frameworks to advance toward a unified handling of uncertainty. Also, there is a need of practical methods to treat information, as using these frameworks can be computationally costly. In this work, we propose some answers to these two needs for a set of commonly encountered problems. In particular, we focus on the problems of: 1) Uncertainty representation 2) Fusion and evaluation of multiple source information 3) Independence modelling, the aim being to give tools (both of theoretical and practical nature) to treat uncertainty. Some tools are then applied to some problems related to nuclear safety issues. (author)

Intelligent Aircraft Damage Assessment, Trajectory Planning, and Decision-Making under Uncertainty

Science.gov (United States)

Lopez, Israel; Sarigul-Klijn, Nesrin

Situational awareness and learning are necessary to identify and select the optimal set of mutually non-exclusive hypothesis in order to maximize mission performance and adapt system behavior accordingly. This paper presents a hierarchical and decentralized approach for integrated damage assessment and trajectory planning in aircraft with uncertain navigational decision-making. Aircraft navigation can be safely accomplished by properly addressing the following: decision-making, obstacle perception, aircraft state estimation, and aircraft control. When in-flight failures or damage occur, rapid and precise decision-making under imprecise information is required in order to regain and maintain control of the aircraft. To achieve planned aircraft trajectory and complete safe landing, the uncertainties in system dynamics of the damaged aircraft need to be learned and incorporated at the level of motion planning. The damaged aircraft is simulated via a simplified kinematic model. The different sources and perspectives of uncertainties in the damage assessment process and post-failure trajectory planning are presented and classified. The decision-making process for an emergency motion planning and landing is developed via the Dempster-Shafer evidence theory. The objective of the trajectory planning is to arrive at a target position while maximizing the safety of the aircraft given uncertain conditions. Simulations are presented for an emergency motion planning and landing that takes into account aircraft dynamics, path complexity, distance to landing site, runway characteristics, and subjective human decision.

HSE's safety assessment principles for criticality safety

International Nuclear Information System (INIS)

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

2008-01-01

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

Avoiding climate change uncertainties in Strategic Environmental Assessment

Energy Technology Data Exchange (ETDEWEB)

Larsen, Sanne Vammen, E-mail: sannevl@plan.aau.dk [The Danish Centre for Environmental Assessment, Aalborg University-Copenhagen, A.C. Meyers Vænge 15, 2450 København SV (Denmark); Kørnøv, Lone, E-mail: lonek@plan.aau.dk [The Danish Centre for Environmental Assessment, Aalborg University, Skibbrogade 5, 1. Sal, 9000 Aalborg (Denmark); Driscoll, Patrick, E-mail: patrick@plan.aau.dk [The Danish Centre for Environmental Assessment, Aalborg University-Copenhagen, A.C. Meyers Vænge 15, 2450 København SV (Denmark)

2013-11-15

This article is concerned with how Strategic Environmental Assessment (SEA) practice handles climate change uncertainties within the Danish planning system. First, a hypothetical model is set up for how uncertainty is handled and not handled in decision-making. The model incorporates the strategies ‘reduction’ and ‘resilience’, ‘denying’, ‘ignoring’ and ‘postponing’. Second, 151 Danish SEAs are analysed with a focus on the extent to which climate change uncertainties are acknowledged and presented, and the empirical findings are discussed in relation to the model. The findings indicate that despite incentives to do so, climate change uncertainties were systematically avoided or downplayed in all but 5 of the 151 SEAs that were reviewed. Finally, two possible explanatory mechanisms are proposed to explain this: conflict avoidance and a need to quantify uncertainty.

Avoiding climate change uncertainties in Strategic Environmental Assessment

International Nuclear Information System (INIS)

Larsen, Sanne Vammen; Kørnøv, Lone; Driscoll, Patrick

2013-01-01

This article is concerned with how Strategic Environmental Assessment (SEA) practice handles climate change uncertainties within the Danish planning system. First, a hypothetical model is set up for how uncertainty is handled and not handled in decision-making. The model incorporates the strategies ‘reduction’ and ‘resilience’, ‘denying’, ‘ignoring’ and ‘postponing’. Second, 151 Danish SEAs are analysed with a focus on the extent to which climate change uncertainties are acknowledged and presented, and the empirical findings are discussed in relation to the model. The findings indicate that despite incentives to do so, climate change uncertainties were systematically avoided or downplayed in all but 5 of the 151 SEAs that were reviewed. Finally, two possible explanatory mechanisms are proposed to explain this: conflict avoidance and a need to quantify uncertainty

Radial core expansion reactivity feedback in advanced LMRs: uncertainties and their effects on inherent safety

International Nuclear Information System (INIS)

Wigeland, R.A.; Moran, T.J.

1988-01-01

An analytical model for calculating radial core expansion, based on the thermal and elastic bowing of a single subassembly at the core periphery, is used to quantify the effect of uncertainties on this reactivity feedback mechanism. This model has been verified and validated with experimental and numerical results. The impact of these uncertainties on the safety margins in unprotected transients is investigated with SASSYS/SAS4A, which includes this model for calculating the reactivity feedback from radial core expansion. The magnitudes of these uncertainties are not sufficient to preclude the use of radial core expansion reactivity feedback in transient analysis

Managing Uncertainties Associated With Radioactive Waste Disposal: Task Group 4 Of The IAEA PRISM Project

International Nuclear Information System (INIS)

Seitz, R.

2011-01-01

It is widely recognized that the results of safety assessment calculations provide an important contribution to the safety arguments for a disposal facility, but cannot in themselves adequately demonstrate the safety of the disposal system. The safety assessment and a broader range of arguments and activities need to be considered holistically to justify radioactive waste disposal at any particular site. Many programs are therefore moving towards the production of what has become known as a Safety Case, which includes all of the different activities that are conducted to demonstrate the safety of a disposal concept. Recognizing the growing interest in the concept of a Safety Case, the International Atomic Energy Agency (IAEA) is undertaking an intercomparison and harmonization project called PRISM (Practical Illustration and use of the Safety Case Concept in the Management of Near-surface Disposal). The PRISM project is organized into four Task Groups that address key aspects of the Safety Case concept: Task Group 1 - Understanding the Safety Case; Task Group 2 - Disposal facility design; Task Group 3 - Managing waste acceptance; and Task Group 4 - Managing uncertainty. This paper addresses the work of Task Group 4, which is investigating approaches for managing the uncertainties associated with near-surface disposal of radioactive waste and their consideration in the context of the Safety Case. Emphasis is placed on identifying a wide variety of approaches that can and have been used to manage different types of uncertainties, especially non-quantitative approaches that have not received as much attention in previous IAEA projects. This paper includes discussions of the current results of work on the task on managing uncertainty, including: the different circumstances being considered, the sources/types of uncertainties being addressed and some initial proposals for approaches that can be used to manage different types of uncertainties.

Uncertainty Evaluation of a Postulated LBLOCA for APR+ using KINS Realistic Evaluation Methodology and MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Hwang, Min Jeong; Marigomena, Ralph; Yoo, Tae Ho; Kim, Y. S.; Sim, S. K. [Environment and Energy Technology, Inc., Daejeon (Korea, Republic of); Bang, Young Seok [KINS, Daejeon (Korea, Republic of)

2014-05-15

As a part of the regulatory safety research, Korea Institute of Nuclear Safety(KINS) also developed a best estimate safety analysis regulatory audit code, MARS-KS, to realistically predict and better understand the physical phenomena of the design basis accidents. KINS improved uncertainty propagation methodology using MARS-KS and applied the improved uncertainty evaluation method for the Shinkori Units 3 and 4 LBLOC. This study is to evaluate the uncertainty propagation of a postulated LBLOCA and quantify the safety margin using KINS-REM and MARS-KS code for the APR+ (Advanced Pressurizer Reactor Plus) Standard Safety Analysis Report(SSAR) which is under regulatory review by the KINS for its design approval. KINS-REM LBLOCA realistic evaluation methodology was used for the regulatory assessment of the APR+ LBLOCA using MARS-KS to evaluate the uncertainty propagation of the uncertainty variables as well as to assess the safety margin during the limiting case of the APR+ double ended guillotine cold leg LBLOCA. Uncertainty evaluation for the APR+ LBLOCA shows that the reflood PCT with upper limit of 95% probability at 95% confidence level is 1363.2 K and is higher than the blowdown PCT95/95 of 1275.3 K. The result shows that the current evaluation of APR+ LBLOCA PCT is within the acceptance criteria of 1477 K ECCS.

Estimation of Uncertainty in Risk Assessment of Hydrogen Applications

DEFF Research Database (Denmark)

Markert, Frank; Krymsky, V.; Kozine, Igor

2011-01-01

Hydrogen technologies such as hydrogen fuelled vehicles and refuelling stations are being tested in practice in a number of projects (e.g. HyFleet-Cute and Whistler project) giving valuable information on the reliability and maintenance requirements. In order to establish refuelling stations the ...... probability and the NUSAP concept to quantify uncertainties of new not fully qualified hydrogen technologies and implications to risk management.......Hydrogen technologies such as hydrogen fuelled vehicles and refuelling stations are being tested in practice in a number of projects (e.g. HyFleet-Cute and Whistler project) giving valuable information on the reliability and maintenance requirements. In order to establish refuelling stations...... the permitting authorities request qualitative and quantitative risk assessments (QRA) to show the safety and acceptability in terms of failure frequencies and respective consequences. For new technologies not all statistical data might be established or are available in good quality causing assumptions...

Safety assessment and verification for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2001-01-01

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

WASH-1400: quantifying the uncertainties

International Nuclear Information System (INIS)

Erdmann, R.C.; Leverenz, F.L. Jr.; Lellouche, G.S.

1981-01-01

The purpose of this paper is to focus on the limitations of the WASH-1400 analysis in estimating the risk from light water reactors (LWRs). This assessment attempts to modify the quantification of the uncertainty in and estimate of risk as presented by the RSS (reactor safety study). 8 refs

Uncertainty of fast biological radiation dose assessment for emergency response scenarios.

Science.gov (United States)

Ainsbury, Elizabeth A; Higueras, Manuel; Puig, Pedro; Einbeck, Jochen; Samaga, Daniel; Barquinero, Joan Francesc; Barrios, Lleonard; Brzozowska, Beata; Fattibene, Paola; Gregoire, Eric; Jaworska, Alicja; Lloyd, David; Oestreicher, Ursula; Romm, Horst; Rothkamm, Kai; Roy, Laurence; Sommer, Sylwester; Terzoudi, Georgia; Thierens, Hubert; Trompier, Francois; Vral, Anne; Woda, Clemens

2017-01-01

Reliable dose estimation is an important factor in appropriate dosimetric triage categorization of exposed individuals to support radiation emergency response. Following work done under the EU FP7 MULTIBIODOSE and RENEB projects, formal methods for defining uncertainties on biological dose estimates are compared using simulated and real data from recent exercises. The results demonstrate that a Bayesian method of uncertainty assessment is the most appropriate, even in the absence of detailed prior information. The relative accuracy and relevance of techniques for calculating uncertainty and combining assay results to produce single dose and uncertainty estimates is further discussed. Finally, it is demonstrated that whatever uncertainty estimation method is employed, ignoring the uncertainty on fast dose assessments can have an important impact on rapid biodosimetric categorization.

Uncertainty evaluation methods for waste package performance assessment

International Nuclear Information System (INIS)

Wu, Y.T.; Nair, P.K.; Journel, A.G.; Abramson, L.R.

1991-01-01

This report identifies and investigates methodologies to deal with uncertainties in assessing high-level nuclear waste package performance. Four uncertainty evaluation methods (probability-distribution approach, bounding approach, expert judgment, and sensitivity analysis) are suggested as the elements of a methodology that, without either diminishing or enhancing the input uncertainties, can evaluate performance uncertainty. Such a methodology can also help identify critical inputs as a guide to reducing uncertainty so as to provide reasonable assurance that the risk objectives are met. This report examines the current qualitative waste containment regulation and shows how, in conjunction with the identified uncertainty evaluation methodology, a framework for a quantitative probability-based rule can be developed that takes account of the uncertainties. Current US Nuclear Regulatory Commission (NRC) regulation requires that the waste packages provide ''substantially complete containment'' (SCC) during the containment period. The term ''SCC'' is ambiguous and subject to interpretation. This report, together with an accompanying report that describes the technical considerations that must be addressed to satisfy high-level waste containment requirements, provides a basis for a third report to develop recommendations for regulatory uncertainty reduction in the ''containment''requirement of 10 CFR Part 60. 25 refs., 3 figs., 2 tabs

Assessment of uncertainties in Neutron Multiplicity Counting

International Nuclear Information System (INIS)

Peerani, P.; Marin Ferrer, M.

2008-01-01

This paper describes a methodology for a complete and correct assessment of the errors coming from the uncertainty of each individual component on the final result. A general methodology accounting for all the main sources of error (both type-A and type-B) will be outlined. In order to better illustrate the method, a practical example applying it to the uncertainty estimation for a special case of multiplicity counter, the SNMC developed at JRC, will be given

Application of intelligence based uncertainty analysis for HLW disposal

International Nuclear Information System (INIS)

Kato, Kazuyuki

2003-01-01

Safety assessment for geological disposal of high level radioactive waste inevitably involves factors that cannot be specified in a deterministic manner. These are namely: (1) 'variability' that arises from stochastic nature of the processes and features considered, e.g., distribution of canister corrosion times and spatial heterogeneity of a host geological formation; (2) 'ignorance' due to incomplete or imprecise knowledge of the processes and conditions expected in the future, e.g., uncertainty in the estimation of solubilities and sorption coefficients for important nuclides. In many cases, a decision in assessment, e.g., selection among model options or determination of a parameter value, is subjected to both variability and ignorance in a combined form. It is clearly important to evaluate both influences of variability and ignorance on the result of a safety assessment in a consistent manner. We developed a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to safety assessment of geological disposal of high level radioactive waste. Uncertainties associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts while variability was formulated by means of probability density functions (pdfs) based on available data set. The exercise demonstrated applicability of the new methodology and, in particular, its advantage in quantifying uncertainties based on expert's opinion and in providing information on dependence of assessment result on the level of conservatism. In addition, it was also shown that sensitivity analysis could identify key parameters in reducing uncertainties associated with the overall assessment. The above information can be used to support the judgment process and guide the process of disposal system development in optimization of protection against

Seismic Hazard Assessment and Uncertainties Treatment: Discussion on the current French regulation, practices and open issues

International Nuclear Information System (INIS)

Berge-Thierry, Catherine

2014-01-01

Taking into account the seismic risk in the context of nuclear safety in France is guided by the Fundamental Safety Rule (RFS2001-01) for the assessment of seismic hazard, and by the Guide ASN/2/01 for the design rules of civil engineering structures. These two references have been updated respectively in 2001 and 2006 and validated by the Nuclear Safety Authority. The French approach is anchored on a deterministic approach. We propose to recall the principles of the methodology recommended by the RFS 2001-01, and to illustrate the advantages and limitations highlighted in recent years. Indeed, this regulatory framework is used both in the design stage and for safety reassessment of all nuclear facilities, power reactors and experimental laboratories and factories. We focus on: (i) key parameters of the approach, and their level of knowledge, (ii) key steps and principles that lead to a non-homogeneous approach between various geographic sites, depending on the seismic activity and / or knowledge, (iii) on physical phenomena (such as the geometric extension of the seismic source, the complexity of earthquake rupture on the fault plane) that are not taken into account, or for which (2D and 3D site effects, and non-linear soil behavior under strong motions), the RFS 2001-01 approach does not provide any guidance, (iv) consideration of epistemic and random uncertainties. We discuss also the probabilistic approaches widely implemented both in France as recently to establish the seismic zoning (reference for the regulation of conventional building and classified installations for the environment), used worldwide and strongly supported by the international Atomic Energy Agency references (safety guides and guidelines). The Tohoku earthquake that occurred in Japan on March 11, 2011, triggering the tsunami that itself caused the nuclear accident at Fukushima Daiichi site has resulted in the realization in France of the Complementary Safety Studies as a request of the

Technology relevance of the 'uncertainty analysis in modelling' project for nuclear reactor safety

International Nuclear Information System (INIS)

D'Auria, F.; Langenbuch, S.; Royer, E.; Del Nevo, A.; Parisi, C.; Petruzzi, A.

2007-01-01

The OECD/NEA Nuclear Science Committee (NSC) endorsed the setting up of an Expert Group on Uncertainty Analysis in Modelling (UAM) in June 2006. This Expert Group reports to the Working Party on Scientific issues in Reactor Systems (WPRS) and because it addresses multi-scale / multi-physics aspects of uncertainty analysis, it will work in close co-ordination with the benchmark groups on coupled neutronics-thermal-hydraulics and on coupled core-plant problems, and the CSNI Group on Analysis and Management of Accidents (GAMA). The NEA/NSC has endorsed that this activity be undertaken with Prof. K. Ivanov from the Pennsylvania State University (PSU) as the main coordinator and host with the assistance of the Scientific Board. The objective of the proposed work is to define, coordinate, conduct, and report an international benchmark for uncertainty analysis in best-estimate coupled code calculations for design, operation, and safety analysis of LWRs entitled 'OECD UAM LWR Benchmark'. At the First Benchmark Workshop (UAM-1) held from 10 to 11 May 2007 at the OECD/NEA, one action concerned the forming of a sub-group, led by F. D'Auria, member of CSNI, responsible for defining the objectives, the impact and benefit of the UAM for safety and licensing. This report is the result of this action by the subgroup. (authors)

Aiding alternatives assessment with an uncertainty-tolerant hazard scoring method.

Science.gov (United States)

Faludi, Jeremy; Hoang, Tina; Gorman, Patrick; Mulvihill, Martin

2016-11-01

This research developed a single-score system to simplify and clarify decision-making in chemical alternatives assessment, accounting for uncertainty. Today, assessing alternatives to hazardous constituent chemicals is a difficult task-rather than comparing alternatives by a single definitive score, many independent toxicological variables must be considered at once, and data gaps are rampant. Thus, most hazard assessments are only comprehensible to toxicologists, but business leaders and politicians need simple scores to make decisions. In addition, they must balance hazard against other considerations, such as product functionality, and they must be aware of the high degrees of uncertainty in chemical hazard data. This research proposes a transparent, reproducible method to translate eighteen hazard endpoints into a simple numeric score with quantified uncertainty, alongside a similar product functionality score, to aid decisions between alternative products. The scoring method uses Clean Production Action's GreenScreen as a guide, but with a different method of score aggregation. It provides finer differentiation between scores than GreenScreen's four-point scale, and it displays uncertainty quantitatively in the final score. Displaying uncertainty also illustrates which alternatives are early in product development versus well-defined commercial products. This paper tested the proposed assessment method through a case study in the building industry, assessing alternatives to spray polyurethane foam insulation containing methylene diphenyl diisocyanate (MDI). The new hazard scoring method successfully identified trade-offs between different alternatives, showing finer resolution than GreenScreen Benchmarking. Sensitivity analysis showed that different weighting schemes in hazard scores had almost no effect on alternatives ranking, compared to uncertainty from data gaps. Copyright © 2016 Elsevier Ltd. All rights reserved.

The role of sensitivity analysis in assessing uncertainty

International Nuclear Information System (INIS)

Crick, M.J.; Hill, M.D.

1987-01-01

Outside the specialist world of those carrying out performance assessments considerable confusion has arisen about the meanings of sensitivity analysis and uncertainty analysis. In this paper we attempt to reduce this confusion. We then go on to review approaches to sensitivity analysis within the context of assessing uncertainty, and to outline the types of test available to identify sensitive parameters, together with their advantages and disadvantages. The views expressed in this paper are those of the authors; they have not been formally endorsed by the National Radiological Protection Board and should not be interpreted as Board advice

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-15

that lie outside the scope of the quantitative safety assessment. These arguments include: Support from natural and anthropogenic analogues for both key process understanding and total system performance. Comparison of the methodology and results with the earlier TILA-99 and SR-Can safety assessments, as well as other international safety assessments, to ensure completeness, consistency and reasonableness of the present assessment. Use of safety indicators other than dose and activity to avoid uncertainties in future human lifestyles and also in geological processes on very long timescales. Consideration of the calculation results from a wider perspective to consider significance of their impact compared to other risks

TREATING UNCERTAINTIES IN A NUCLEAR SEISMIC PROBABILISTIC RISK ASSESSMENT BY MEANS OF THE DEMPSTER-SHAFER THEORY OF EVIDENCE

Directory of Open Access Journals (Sweden)

CHUNG-KUNG LO

2014-02-01

Full Text Available The analyses carried out within the Seismic Probabilistic Risk Assessments (SPRAs of Nuclear Power Plants (NPPs are affected by significant aleatory and epistemic uncertainties. These uncertainties have to be represented and quantified coherently with the data, information and knowledge available, to provide reasonable assurance that related decisions can be taken robustly and with confidence. The amount of data, information and knowledge available for seismic risk assessment is typically limited, so that the analysis must strongly rely on expert judgments. In this paper, a Dempster-Shafer Theory (DST framework for handling uncertainties in NPP SPRAs is proposed and applied to an example case study. The main contributions of this paper are two: (i applying the complete DST framework to SPRA models, showing how to build the Dempster-Shafer structures of the uncertainty parameters based on industry generic data, and (ii embedding Bayesian updating based on plant specific data into the framework. The results of the application to a case study show that the approach is feasible and effective in (i describing and jointly propagating aleatory and epistemic uncertainties in SPRA models and (ii providing ‘conservative’ bounds on the safety quantities of interest (i.e. Core Damage Frequency, CDF that reflect the (limited state of knowledge of the experts about the system of interest.

A review of uncertainty research in impact assessment

International Nuclear Information System (INIS)

Leung, Wanda; Noble, Bram; Gunn, Jill; Jaeger, Jochen A.G.

2015-01-01

This paper examines uncertainty research in Impact Assessment (IA) and the focus of attention of the IA scholarly literature. We do so by first exploring ‘outside’ the IA literature, identifying three main themes of uncertainty research, and then apply these themes to examine the focus of scholarly research on uncertainty ‘inside’ IA. Based on a search of the database Scopus, we identified 134 journal papers published between 1970 and 2013 that address uncertainty in IA, 75% of which were published since 2005. We found that 90% of IA research addressing uncertainty focused on uncertainty in the practice of IA, including uncertainty in impact predictions, models and managing environmental impacts. Notwithstanding early guidance on uncertainty treatment in IA from the 1980s, we found no common, underlying conceptual framework that was guiding research on uncertainty in IA practice. Considerably less attention, only 9% of papers, focused on uncertainty communication, disclosure and decision-making under uncertain conditions, the majority of which focused on the need to disclose uncertainties as opposed to providing guidance on how to do so and effectively use that information to inform decisions. Finally, research focused on theory building for explaining human behavior with respect to uncertainty avoidance constituted only 1% of the IA published literature. We suggest the need for further conceptual framework development for researchers focused on identifying and addressing uncertainty in IA practice; the need for guidance on how best to communicate uncertainties in practice, versus criticizing practitioners for not doing so; research that explores how best to interpret and use disclosures about uncertainty when making decisions about project approvals, and the implications of doing so; and academic theory building and exploring the utility of existing theories to better understand and explain uncertainty avoidance behavior in IA. - Highlights: • We

A review of uncertainty research in impact assessment

Energy Technology Data Exchange (ETDEWEB)

Leung, Wanda, E-mail: wanda.leung@usask.ca [Department of Geography and Planning, University of Saskatchewan, 117 Science Place, Saskatoon, Saskatchewan S7N 5A5 (Canada); Noble, Bram, E-mail: b.noble@usask.ca [Department of Geography and Planning, School of Environment and Sustainability, University of Saskatchewan, 117 Science Place, Saskatoon, Saskatchewan S7N 5A5 (Canada); Gunn, Jill, E-mail: jill.gunn@usask.ca [Department of Geography and Planning, University of Saskatchewan, 117 Science Place, Saskatoon, Saskatchewan S7N 5A5 (Canada); Jaeger, Jochen A.G., E-mail: jochen.jaeger@concordia.ca [Department of Geography, Planning and Environment, Concordia University, 1455 de Maisonneuve W., Suite 1255, Montreal, Quebec H3G 1M8 (Canada); Loyola Sustainability Research Centre, Concordia University, 7141 Sherbrooke W., AD-502, Montreal, Quebec H4B 1R6 (Canada)

2015-01-15

This paper examines uncertainty research in Impact Assessment (IA) and the focus of attention of the IA scholarly literature. We do so by first exploring ‘outside’ the IA literature, identifying three main themes of uncertainty research, and then apply these themes to examine the focus of scholarly research on uncertainty ‘inside’ IA. Based on a search of the database Scopus, we identified 134 journal papers published between 1970 and 2013 that address uncertainty in IA, 75% of which were published since 2005. We found that 90% of IA research addressing uncertainty focused on uncertainty in the practice of IA, including uncertainty in impact predictions, models and managing environmental impacts. Notwithstanding early guidance on uncertainty treatment in IA from the 1980s, we found no common, underlying conceptual framework that was guiding research on uncertainty in IA practice. Considerably less attention, only 9% of papers, focused on uncertainty communication, disclosure and decision-making under uncertain conditions, the majority of which focused on the need to disclose uncertainties as opposed to providing guidance on how to do so and effectively use that information to inform decisions. Finally, research focused on theory building for explaining human behavior with respect to uncertainty avoidance constituted only 1% of the IA published literature. We suggest the need for further conceptual framework development for researchers focused on identifying and addressing uncertainty in IA practice; the need for guidance on how best to communicate uncertainties in practice, versus criticizing practitioners for not doing so; research that explores how best to interpret and use disclosures about uncertainty when making decisions about project approvals, and the implications of doing so; and academic theory building and exploring the utility of existing theories to better understand and explain uncertainty avoidance behavior in IA. - Highlights: • We

Assessment and uncertainty analysis of groundwater risk.

Science.gov (United States)

Li, Fawen; Zhu, Jingzhao; Deng, Xiyuan; Zhao, Yong; Li, Shaofei

2018-01-01

Groundwater with relatively stable quantity and quality is commonly used by human being. However, as the over-mining of groundwater, problems such as groundwater funnel, land subsidence and salt water intrusion have emerged. In order to avoid further deterioration of hydrogeological problems in over-mining regions, it is necessary to conduct the assessment of groundwater risk. In this paper, risks of shallow and deep groundwater in the water intake area of the South-to-North Water Transfer Project in Tianjin, China, were evaluated. Firstly, two sets of four-level evaluation index system were constructed based on the different characteristics of shallow and deep groundwater. Secondly, based on the normalized factor values and the synthetic weights, the risk values of shallow and deep groundwater were calculated. Lastly, the uncertainty of groundwater risk assessment was analyzed by indicator kriging method. The results meet the decision maker's demand for risk information, and overcome previous risk assessment results expressed in the form of deterministic point estimations, which ignore the uncertainty of risk assessment. Copyright © 2017 Elsevier Inc. All rights reserved.

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

Uncertainty and Risk Assessment in the Design Process for Wind

Energy Technology Data Exchange (ETDEWEB)

Damiani, Rick R. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2018-02-09

This report summarizes the concepts and opinions that emerged from an initial study on the subject of uncertainty in wind design that included expert elicitation during a workshop held at the National Wind Technology Center at the National Renewable Energy Laboratory July 12-13, 2016. In this paper, five major categories of uncertainties are identified. The first category is associated with direct impacts on turbine loads, (i.e., the inflow including extreme events, aero-hydro-servo-elastic response, soil-structure inter- action, and load extrapolation). The second category encompasses material behavior and strength. Site suitability and due-diligence aspects pertain to the third category. Calibration of partial safety factors and optimal reliability levels make up the fourth one. And last but not least, is the category associated with uncertainties in computational modeling. The main sections of this paper follow this organization.

Expert judgment based multi-criteria decision model to address uncertainties in risk assessment of nanotechnology-enabled food products

International Nuclear Information System (INIS)

Flari, Villie; Chaudhry, Qasim; Neslo, Rabin; Cooke, Roger

2011-01-01

Currently, risk assessment of nanotechnology-enabled food products is considered difficult due to the large number of uncertainties involved. We developed an approach which could address some of the main uncertainties through the use of expert judgment. Our approach employs a multi-criteria decision model, based on probabilistic inversion that enables capturing experts’ preferences in regard to safety of nanotechnology-enabled food products, and identifying their opinions in regard to the significance of key criteria that are important in determining the safety of such products. An advantage of these sample-based techniques is that they provide out-of-sample validation and therefore a robust scientific basis. This validation in turn adds predictive power to the model developed. We achieved out-of-sample validation in two ways: (1) a portion of the expert preference data was excluded from the model’s fitting and was then predicted by the model fitted on the remaining rankings and (2) a (partially) different set of experts generated new scenarios, using the same criteria employed in the model, and ranked them; their ranks were compared with ranks predicted by the model. The degree of validation in each method was less than perfect but reasonably substantial. The validated model we applied captured and modelled experts’ preferences regarding safety of hypothetical nanotechnology-enabled food products. It appears therefore that such an approach can provide a promising route to explore further for assessing the risk of nanotechnology-enabled food products.

The Uncertainty Test for the MAAP Computer Code

International Nuclear Information System (INIS)

Park, S. H.; Song, Y. M.; Park, S. Y.; Ahn, K. I.; Kim, K. R.; Lee, Y. J.

2008-01-01

After the Three Mile Island Unit 2 (TMI-2) and Chernobyl accidents, safety issues for a severe accident are treated in various aspects. Major issues in our research part include a level 2 PSA. The difficulty in expanding the level 2 PSA as a risk information activity is the uncertainty. In former days, it attached a weight to improve the quality in a internal accident PSA, but the effort is insufficient for decrease the phenomenon uncertainty in the level 2 PSA. In our country, the uncertainty degree is high in the case of a level 2 PSA model, and it is necessary to secure a model to decrease the uncertainty. We have not yet experienced the uncertainty assessment technology, the assessment system itself depends on advanced nations. In advanced nations, the severe accident simulator is implemented in the hardware level. But in our case, basic function in a software level can be implemented. In these circumstance at home and abroad, similar instances are surveyed such as UQM and MELCOR. Referred to these instances, SAUNA (Severe Accident UNcertainty Analysis) system is being developed in our project to assess and decrease the uncertainty in a level 2 PSA. It selects the MAAP code to analyze the uncertainty in a severe accident

Development of reliability and probabilistic safety assessment program RiskA

International Nuclear Information System (INIS)

Wu, Yican

2015-01-01

Highlights: • There are four parts in the structure of RiskA. User input part lets users input the PSA model and some necessary data by GUI or model transformation tool. In calculation engine part, fault tree analysis, event tree analysis, uncertainty analysis, sensitivity analysis, importance analysis and failure mode and effects analysis are supplied. User output part outputs the analysis results, user customized reports and some other data. The last part includes reliability database, some other common tools and help documents. • RiskA has several advanced features. Extensible framework makes it easy to add any new functions, making RiskA to be a large platform of reliability and probabilistic safety assessment. It is very fast to analysis fault tree in RiskA because many advanced algorithm improvement were made. Many model formats can be imported and exported, which made the PSA model in the commercial software can be easily transformed to adapt RiskA platform. Web-based co-modeling let several users in different places work together whenever they are online. • The comparison between RiskA and other mature PSA codes (e.g. CAFTA, RiskSpectrum, XFTA) has demonstrated that the calculation and analysis of RiskA is correct and efficient. Based on the development of this code package, many applications of safety and reliability analysis of some research reactors and nuclear power plants were performed. The development of RiskA appears to be of realistic and potential value for academic research and practical operation safety management of nuclear power plants in China and abroad. - Abstract: PSA (probabilistic safety assessment) software, the indispensable tool in nuclear safety assessment, has been widely used. An integrated reliability and PSA program named RiskA has been developed by FDS Team. RiskA supplies several standard PSA modules including fault tree analysis, event tree analysis, uncertainty analysis, failure mode and effect analysis and reliability

Structural observation of long-span suspension bridges for safety assessment: implementation of an optical displacement measurement system

International Nuclear Information System (INIS)

Martins, L Lages; Ribeiro, A Silva; Rebordão, J M

2015-01-01

This paper addresses the implementation of an optical displacement measurement system in the observation scenario of a long-span suspension bridge and its contribution for structural safety assessment. The metrological background required for quality assurance of the measurements is described, namely, the system's intrinsic parameterization and integration in the SI dimensional traceability chain by calibration, including its measurement uncertainty assessment

International survey for good practices in forecasting uncertainty assessment and communication

Science.gov (United States)

Berthet, Lionel; Piotte, Olivier

2014-05-01

Achieving technically sound flood forecasts is a crucial objective for forecasters but remains of poor use if the users do not understand properly their significance and do not use it properly in decision making. One usual way to precise the forecasts limitations is to communicate some information about their uncertainty. Uncertainty assessment and communication to stakeholders are thus important issues for operational flood forecasting services (FFS) but remain open fields for research. French FFS wants to publish graphical streamflow and level forecasts along with uncertainty assessment in near future on its website (available to the greater public). In order to choose the technical options best adapted to its operational context, it carried out a survey among more than 15 fellow institutions. Most of these are providing forecasts and warnings to civil protection officers while some were mostly working for hydroelectricity suppliers. A questionnaire has been prepared in order to standardize the analysis of the practices of the surveyed institutions. The survey was conducted by gathering information from technical reports or from the scientific literature, as well as 'interviews' driven by phone, email discussions or meetings. The questionnaire helped in the exploration of practices in uncertainty assessment, evaluation and communication. Attention was paid to the particular context within which every insitution works, in the analysis drawn from raw results. Results show that most services interviewed assess their forecasts uncertainty. However, practices can differ significantly from a country to another. Popular techniques are ensemble approaches. They allow to take into account several uncertainty sources. Statistical past forecasts analysis (such as the quantile regressions) are also commonly used. Contrary to what was expected, only few services emphasize the role of the forecaster (subjective assessment). Similar contrasts can be observed in uncertainty

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

An introductory guide to uncertainty analysis in environmental and health risk assessment

International Nuclear Information System (INIS)

Hoffman, F.O.; Hammonds, J.S.

1992-10-01

To compensate for the potential for overly conservative estimates of risk using standard US Environmental Protection Agency methods, an uncertainty analysis should be performed as an integral part of each risk assessment. Uncertainty analyses allow one to obtain quantitative results in the form of confidence intervals that will aid in decision making and will provide guidance for the acquisition of additional data. To perform an uncertainty analysis, one must frequently rely on subjective judgment in the absence of data to estimate the range and a probability distribution describing the extent of uncertainty about a true but unknown value for each parameter of interest. This information is formulated from professional judgment based on an extensive review of literature, analysis of the data, and interviews with experts. Various analytical and numerical techniques are available to allow statistical propagation of the uncertainty in the model parameters to a statement of uncertainty in the risk to a potentially exposed individual. Although analytical methods may be straightforward for relatively simple models, they rapidly become complicated for more involved risk assessments. Because of the tedious efforts required to mathematically derive analytical approaches to propagate uncertainty in complicated risk assessments, numerical methods such as Monte Carlo simulation should be employed. The primary objective of this report is to provide an introductory guide for performing uncertainty analysis in risk assessments being performed for Superfund sites

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

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1984-08-01

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

Decay heat uncertainty quantification of MYRRHA

OpenAIRE

Fiorito Luca; Buss Oliver; Hoefer Axel; Stankovskiy Alexey; Eynde Gert Van den

2017-01-01

MYRRHA is a lead-bismuth cooled MOX-fueled accelerator driven system (ADS) currently in the design phase at SCK·CEN in Belgium. The correct evaluation of the decay heat and of its uncertainty level is very important for the safety demonstration of the reactor. In the first part of this work we assessed the decay heat released by the MYRRHA core using the ALEPH-2 burnup code. The second part of the study focused on the nuclear data uncertainty and covariance propagation to the MYRRHA decay hea...

Problems of probabilistic safety assessment after Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

Sugiyama, Naoki

2011-01-01

Probabilistic safety assessment (PSA) methodology to assure nuclear safety is had great expectations of lessons learned from Fukushima Daiichi nuclear power plant (NPP) accident and on the other hand this accident made actualized technical problems of PSA. Effectiveness of current PSA methodology for risk assessment was confirmed by comparing the accident development with accident scenario of PSA and equipment failure rate. From a viewpoint of nuclear safety objective and defense in depth approach of IAEA, technical problems of PSA were (1) extension of PSA for spent fuel pool and waste disposal system as well as level 3PSA for broader environmental contamination and (2) overlapping of accident scenario of plural unit site, balance of high quality plant management and preceding negation, treatment of uncertainty of external events, severe accident measure and human reliability analysis and reflection of disaster prevention capability to level 3PSA. In order to upgrade PSA technology, six proposals were described for nuclear safety and defense in depth, comprehensive evaluation scope and catch-up of latest technology, necessity of strategic preparation of PSA standard, human resources fostering and risk communication. (T. Tanaka)

Joint analysis of epistemic and aleatory uncertainty in stability analysis for geo-hazard assessments

Science.gov (United States)

Rohmer, Jeremy; Verdel, Thierry

2017-04-01

Uncertainty analysis is an unavoidable task of stability analysis of any geotechnical systems. Such analysis usually relies on the safety factor SF (if SF is below some specified threshold), the failure is possible). The objective of the stability analysis is then to estimate the failure probability P for SF to be below the specified threshold. When dealing with uncertainties, two facets should be considered as outlined by several authors in the domain of geotechnics, namely "aleatoric uncertainty" (also named "randomness" or "intrinsic variability") and "epistemic uncertainty" (i.e. when facing "vague, incomplete or imprecise information" such as limited databases and observations or "imperfect" modelling). The benefits of separating both facets of uncertainty can be seen from a risk management perspective because: - Aleatoric uncertainty, being a property of the system under study, cannot be reduced. However, practical actions can be taken to circumvent the potentially dangerous effects of such variability; - Epistemic uncertainty, being due to the incomplete/imprecise nature of available information, can be reduced by e.g., increasing the number of tests (lab or in site survey), improving the measurement methods or evaluating calculation procedure with model tests, confronting more information sources (expert opinions, data from literature, etc.). Uncertainty treatment in stability analysis usually restricts to the probabilistic framework to represent both facets of uncertainty. Yet, in the domain of geo-hazard assessments (like landslides, mine pillar collapse, rockfalls, etc.), the validity of this approach can be debatable. In the present communication, we propose to review the major criticisms available in the literature against the systematic use of probability in situations of high degree of uncertainty. On this basis, the feasibility of using a more flexible uncertainty representation tool is then investigated, namely Possibility distributions (e

A generalized framework for assessment of safety margins in nuclear power plants

International Nuclear Information System (INIS)

Gavrilas, M.; Youngblood, B.; Prelewicz, D.; Meyer, Jim

2004-01-01

The protection of public health and safety, and the environment from inadvertent releases of radioactive materials from nuclear power plants relies on the implementation of the defense-in-depth strategy. The term defense-in-depth evolved historically, and thus its application has not always been uniform. The use of the term in the context of the U.S. Nuclear Regulatory Commission (NRC) safety philosophy entails the reliance of a nuclear facility on successive compensatory measures in preventing accidents or mitigating damage caused by malfunctions, accidents, or naturally occurring events. The introduction of probabilistic risk analyses with NUREG-74/014 and subsequent evolution in risk assessment techniques, are leading to the implementation of risk informed regulation to ensure the safety of the public and the environment. Risk informed regulation minimizes the likelihood of overlooking potentially significant accident sequences while limiting unnecessary burdens imposed on licensees. The proposed framework merges fundamental elements of safety regulation: defense-in depth, safety margins and probabilistic risk. It formalizes the relationship between probabilistic risk assessment (PRA) methods and data, and deterministic analyses in a manner consistent with NRC's defense-in-depth philosophy. Succinctly put, the likelihood and consequences of accident scenarios are considered simultaneously and quantified by a plant safety metric. The integration of these fundamental elements into a practically applicable safety framework is consistent with the NRC policy statement on use of probabilistic risk assessment methods and the November 2002 Regulatory Guide on risk informed decisions on plant-specific changes to the licensing basis. Safety information resulting from the application of the framework supersedes traditional safety figures of merit. Safety quantifiers, referred herein as safety indices, expand on the qualifier outcomes that currently accompany fault tree

Sources/treatment of uncertainties in the performance assessment of geologic radioactive waste repositories

International Nuclear Information System (INIS)

Cranwell, R.M.

1987-01-01

Uncertainties in the performance assessment of geologic radioactive waste repositories have several sources. The more important ones include: 1) uncertainty in the conditions of a disposal system over the temporal scales set forth in regulations, 2) uncertainty in the conceptualization of the geohydrologic system, 3) uncertainty in the theoretical description of a given conceptual model of the system, 4) uncertainty in the development of computer codes to implement the solution of a mathematical model, and 5) uncertainty in the parameters and data required in the models and codes used to assess the long-term performance of the disposal system. This paper discusses each of these uncertainties and outlines methods for addressing these uncertainties

Approaches for treating uncertainty in the long term performance assessment of a geological waste repository in clay

International Nuclear Information System (INIS)

Marivoet, J.; Volckaert, G.; Wemaere, I.; Mallants, D.

1998-01-01

In Belgium the current strategy for high-level waste disposal is the geological disposal in a plastic clay layer. The performance assessment approach consists of a systematic scenario selection based on the FEP (features, events and processes) methodology followed by consequence analyses for the selected scenarios. In these consequence analyses the different sources of uncertainty are systematically considered. For the normal evolution scenario, i.e. the scenario which includes all FEP's which are about certain to occur, a stochastic technique of the Monte Carlo type is applied for treating uncertainty. For the altered evolution scenarios a deterministic approach is generally used to evaluate the uncertainties on the long-term. In the case of altered evolution scenarios, comparisons of fluxes from the far field into the biosphere with those calculated for the normal evolution scenario are used, beside dose calculations, to evaluate the safety consequences. Some typical examples of the above approaches will be presented. (author)

Assessment of Measurement Uncertainty Values of the Scandium Determination in Marine Sediment

International Nuclear Information System (INIS)

Rina-Mulyaningsih, Th.

2005-01-01

The result value of testing is meaningless if it isn't completed with uncertainty value. So that with the analysis result Sc in the marine sediment sample. It was assessed the uncertainty measurement of Sc analysis in marine sediment. The experiment was done in AAN Serpong laboratory. The result of calculation uncertainty on Sc analysis showed that the uncertainty components come from: preparation of sample and standard/comparator, purity of standard, counting statistics (sample and standard), repeatability, nuclear data and decay correction. The assessment on uncertainty must be done for the analysis of others elements, because each elements has difference nuclear and physical properties. (author)

Error and Uncertainty in the Accuracy Assessment of Land Cover Maps

Science.gov (United States)

Sarmento, Pedro Alexandre Reis

Traditionally the accuracy assessment of land cover maps is performed through the comparison of these maps with a reference database, which is intended to represent the "real" land cover, being this comparison reported with the thematic accuracy measures through confusion matrixes. Although, these reference databases are also a representation of reality, containing errors due to the human uncertainty in the assignment of the land cover class that best characterizes a certain area, causing bias in the thematic accuracy measures that are reported to the end users of these maps. The main goal of this dissertation is to develop a methodology that allows the integration of human uncertainty present in reference databases in the accuracy assessment of land cover maps, and analyse the impacts that uncertainty may have in the thematic accuracy measures reported to the end users of land cover maps. The utility of the inclusion of human uncertainty in the accuracy assessment of land cover maps is investigated. Specifically we studied the utility of fuzzy sets theory, more precisely of fuzzy arithmetic, for a better understanding of human uncertainty associated to the elaboration of reference databases, and their impacts in the thematic accuracy measures that are derived from confusion matrixes. For this purpose linguistic values transformed in fuzzy intervals that address the uncertainty in the elaboration of reference databases were used to compute fuzzy confusion matrixes. The proposed methodology is illustrated using a case study in which the accuracy assessment of a land cover map for Continental Portugal derived from Medium Resolution Imaging Spectrometer (MERIS) is made. The obtained results demonstrate that the inclusion of human uncertainty in reference databases provides much more information about the quality of land cover maps, when compared with the traditional approach of accuracy assessment of land cover maps. None

Uncertainty analysis in Monte Carlo criticality computations

International Nuclear Information System (INIS)

Qi Ao

2011-01-01

Highlights: ► Two types of uncertainty methods for k eff Monte Carlo computations are examined. ► Sampling method has the least restrictions on perturbation but computing resources. ► Analytical method is limited to small perturbation on material properties. ► Practicality relies on efficiency, multiparameter applicability and data availability. - Abstract: Uncertainty analysis is imperative for nuclear criticality risk assessments when using Monte Carlo neutron transport methods to predict the effective neutron multiplication factor (k eff ) for fissionable material systems. For the validation of Monte Carlo codes for criticality computations against benchmark experiments, code accuracy and precision are measured by both the computational bias and uncertainty in the bias. The uncertainty in the bias accounts for known or quantified experimental, computational and model uncertainties. For the application of Monte Carlo codes for criticality analysis of fissionable material systems, an administrative margin of subcriticality must be imposed to provide additional assurance of subcriticality for any unknown or unquantified uncertainties. Because of a substantial impact of the administrative margin of subcriticality on economics and safety of nuclear fuel cycle operations, recently increasing interests in reducing the administrative margin of subcriticality make the uncertainty analysis in criticality safety computations more risk-significant. This paper provides an overview of two most popular k eff uncertainty analysis methods for Monte Carlo criticality computations: (1) sampling-based methods, and (2) analytical methods. Examples are given to demonstrate their usage in the k eff uncertainty analysis due to uncertainties in both neutronic and non-neutronic parameters of fissionable material systems.

Methodology for qualitative uncertainty assessment of climate impact indicators

Science.gov (United States)

Otto, Juliane; Keup-Thiel, Elke; Rechid, Diana; Hänsler, Andreas; Pfeifer, Susanne; Roth, Ellinor; Jacob, Daniela

2016-04-01

The FP7 project "Climate Information Portal for Copernicus" (CLIPC) is developing an integrated platform of climate data services to provide a single point of access for authoritative scientific information on climate change and climate change impacts. In this project, the Climate Service Center Germany (GERICS) has been in charge of the development of a methodology on how to assess the uncertainties related to climate impact indicators. Existing climate data portals mainly treat the uncertainties in two ways: Either they provide generic guidance and/or express with statistical measures the quantifiable fraction of the uncertainty. However, none of the climate data portals give the users a qualitative guidance how confident they can be in the validity of the displayed data. The need for such guidance was identified in CLIPC user consultations. Therefore, we aim to provide an uncertainty assessment that provides the users with climate impact indicator-specific guidance on the degree to which they can trust the outcome. We will present an approach that provides information on the importance of different sources of uncertainties associated with a specific climate impact indicator and how these sources affect the overall 'degree of confidence' of this respective indicator. To meet users requirements in the effective communication of uncertainties, their feedback has been involved during the development process of the methodology. Assessing and visualising the quantitative component of uncertainty is part of the qualitative guidance. As visual analysis method, we apply the Climate Signal Maps (Pfeifer et al. 2015), which highlight only those areas with robust climate change signals. Here, robustness is defined as a combination of model agreement and the significance of the individual model projections. Reference Pfeifer, S., Bülow, K., Gobiet, A., Hänsler, A., Mudelsee, M., Otto, J., Rechid, D., Teichmann, C. and Jacob, D.: Robustness of Ensemble Climate Projections

Assessment of Safety Culture

International Nuclear Information System (INIS)

Bilic Zabric, T.; Kavsek, D.

2006-01-01

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

Assessment of dose measurement uncertainty using RisoScan

International Nuclear Information System (INIS)

Helt-Hansen, Jakob; Miller, Arne

2006-01-01

The dose measurement uncertainty of the dosimeter system RisoScan, office scanner and Riso B3 dosimeters has been assessed by comparison with spectrophotometer measurements of the same dosimeters. The reproducibility and the combined uncertainty were found to be approximately 2% and 4%, respectively, at one standard deviation. The subroutine in RisoScan for electron energy measurement is shown to give results that are equivalent to the measurements with a scanning spectrophotometer

Light water reactor sequence timing: its significance to probabilistic safety assessment modeling

International Nuclear Information System (INIS)

Bley, D.C.; Buttemer, D.R.; Stetkar, J.W.

1988-01-01

This paper examines event sequence timing in light water reactor plants from the viewpoint of probabilistic safety assessment (PSA). The analytical basis for the ideas presented here come primarily from the authors' work in support of more than 20 PSA studies over the past several years. Timing effects are important for establishing success criteria for support and safety system response and for identifying the time available for operator recovery actions. The principal results of this paper are as follows: 1. Analysis of event sequence timing is necessary for meaningful probabilistic safety assessment - both the success criteria for systems performance and the probability of recovery are tightly linked to sequence timing. 2. Simple engineering analyses based on first principles are often sufficient to provide adequate resolution of the time available for recovery of PSA scenarios. Only those parameters that influence sequence timing and its variability and uncertainty need be examined. 3. Time available for recovery is the basic criterion for evaluation of human performance, whether time is an explicit parameter of the operator actions analysis or not. (author)

Quantifying uncertainty and trade-offs in resilience assessments

Directory of Open Access Journals (Sweden)

Craig R. Allen

2018-03-01

Full Text Available Several frameworks have been developed to assess the resilience of social-ecological systems, but most require substantial data inputs, time, and technical expertise. Stakeholders and practitioners often lack the resources for such intensive efforts. Furthermore, most end with problem framing and fail to explicitly address trade-offs and uncertainty. To remedy this gap, we developed a rapid survey assessment that compares the relative resilience of social-ecological systems with respect to a number of resilience properties. This approach generates large amounts of information relative to stakeholder inputs. We targeted four stakeholder categories: government (policy, regulation, management, end users (farmers, ranchers, landowners, industry, agency/public science (research, university, extension, and NGOs (environmental, citizen, social justice in four North American watersheds, to assess social-ecological resilience through surveys. Conceptually, social-ecological systems are comprised of components ranging from strictly human to strictly ecological, but that relate directly or indirectly to one another. They have soft boundaries and several important dimensions or axes that together describe the nature of social-ecological interactions, e.g., variability, diversity, modularity, slow variables, feedbacks, capital, innovation, redundancy, and ecosystem services. There is no absolute measure of resilience, so our design takes advantage of cross-watershed comparisons and therefore focuses on relative resilience. Our approach quantifies and compares the relative resilience across watershed systems and potential trade-offs among different aspects of the social-ecological system, e.g., between social, economic, and ecological contributions. This approach permits explicit assessment of several types of uncertainty (e.g., self-assigned uncertainty for stakeholders; uncertainty across respondents, watersheds, and subsystems, and subjectivity in

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

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

Uncertainties in environmental radiological assessment models and their implications

International Nuclear Information System (INIS)

Hoffman, F.O.; Miller, C.W.

1983-01-01

Environmental radiological assessments rely heavily on the use of mathematical models. The predictions of these models are inherently uncertain because these models are inexact representations of real systems. The major sources of this uncertainty are related to biases in model formulation and parameter estimation. The best approach for estimating the actual extent of over- or underprediction is model validation, a procedure that requires testing over the range of the intended realm of model application. Other approaches discussed are the use of screening procedures, sensitivity and stochastic analyses, and model comparison. The magnitude of uncertainty in model predictions is a function of the questions asked of the model and the specific radionuclides and exposure pathways of dominant importance. Estimates are made of the relative magnitude of uncertainty for situations requiring predictions of individual and collective risks for both chronic and acute releases of radionuclides. It is concluded that models developed as research tools should be distinguished from models developed for assessment applications. Furthermore, increased model complexity does not necessarily guarantee increased accuracy. To improve the realism of assessment modeling, stochastic procedures are recommended that translate uncertain parameter estimates into a distribution of predicted values. These procedures also permit the importance of model parameters to be ranked according to their relative contribution to the overall predicted uncertainty. Although confidence in model predictions can be improved through site-specific parameter estimation and increased model validation, risk factors and internal dosimetry models will probably remain important contributors to the amount of uncertainty that is irreducible

A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Aldemir, Tunc

2013-01-01

Highlights: ► Dynamic methodologies for probabilistic safety assessment (PSA) are surveyed. ► These methodologies overcome the limitations of the traditional approach to PSA. ► They are suitable for PSA using a best estimate plus uncertainty approach. ► They are highly computation intensive and produce very large number of scenarios. ► Use of scenario clustering can assist the analysis of the results. -- Abstract: Dynamic methodologies for probabilistic safety assessment (PSA) are defined as those which use a time-dependent phenomenological model of system evolution along with its stochastic behavior to account for possible dependencies between failure events. Over the past 30 years, numerous concerns have been raised in the literature regarding the capability of the traditional static modeling approaches such as the event-tree/fault-tree methodology to adequately account for the impact of process/hardware/software/firmware/human interactions on the stochastic system behavior. A survey of the types of dynamic PSA methodologies proposed to date is presented, as well as a brief summary of an example application for the PSA modeling of a digital feedwater control system of an operating pressurized water reactor. The use of dynamic methodologies for PSA modeling of passive components and phenomenological uncertainties are also discussed.

Uncertainty assessment for accelerator-driven systems

International Nuclear Information System (INIS)

Finck, P. J.; Gomes, I.; Micklich, B.; Palmiotti, G.

1999-01-01

The concept of a subcritical system driven by an external source of neutrons provided by an accelerator ADS (Accelerator Driver System) has been recently revived and is becoming more popular in the world technical community with active programs in Europe, Russia, Japan, and the U.S. A general consensus has been reached in adopting for the subcritical component a fast spectrum liquid metal cooled configuration. Both a lead-bismuth eutectic, sodium and gas are being considered as a coolant; each has advantages and disadvantages. The major expected advantage is that subcriticality avoids reactivity induced transients. The potentially large subcriticality margin also should allow for the introduction of very significant quantities of waste products (minor Actinides and Fission Products) which negatively impact the safety characteristics of standard cores. In the U.S. these arguments are the basis for the development of the Accelerator Transmutation of Waste (ATW), which has significant potential in reducing nuclear waste levels. Up to now, neutronic calculations have not attached uncertainties on the values of the main nuclear integral parameters that characterize the system. Many of these parameters (e.g., degree of subcriticality) are crucial to demonstrate the validity and feasibility of this concept. In this paper we will consider uncertainties related to nuclear data only. The present knowledge of the cross sections of many isotopes that are not usually utilized in existing reactors (like Bi, Pb-207, Pb-208, and also Minor Actinides and Fission Products) suggests that uncertainties in the integral parameters will be significantly larger than for conventional reactor systems, and this raises concerns on the neutronic performance of those systems

Uncertainty of Energy Consumption Assessment of Domestic Buildings

DEFF Research Database (Denmark)

Brohus, Henrik; Heiselberg, Per; Simonsen, A.

2009-01-01

In order to assess the influence of energy reduction initiatives, to determine the expected annual cost, to calculate life cycle cost, emission impact, etc. it is crucial to be able to assess the energy consumption reasonably accurate. The present work undertakes a theoretical and empirical study...... of the uncertainty of energy consumption assessment of domestic buildings. The calculated energy consumption of a number of almost identical domestic buildings in Denmark is compared with the measured energy consumption. Furthermore, the uncertainty is determined by means of stochastic modelling based on input...... to correspond reasonably well; however, it is also found that significant differences may occur between calculated and measured energy consumption due to the spread and due to the fact that the result can only be determined with a certain probability. It is found that occupants' behaviour is the major...

Radioactive waste disposal system for Cuba. Safety assessment for the long term

International Nuclear Information System (INIS)

Peralta Vital, J.L.; Gil Castillo, R.; Mirta Torrez, B.

1998-01-01

The present work is performed within the frame of evaluating the radiological impact of the post-closure stage of the facility for disposal of the radioactive wastes generated in Cuba, including a description of the waste disposal systems defined in the country, and taking account of significant elements of their long term safety. The Methodology for Safety Assessment includes: the definition of possible scenarios for evaluation, the identification of principal present uncertainties, the model simulating the release of the radionuclides of the facility, their transport through the geosphere, and their final access to man, evaluating ultimately the radiological impact of the disposal system considering the dose for a critical group. The results obtained allow to demonstrate the radiological safety of the nominative barrier in the design of the system for the particular conditions of Cuba. (author)

Evaluating variability and uncertainty in radiological impact assessment using SYMBIOSE

International Nuclear Information System (INIS)

Simon-Cornu, M.; Beaugelin-Seiller, K.; Boyer, P.; Calmon, P.; Garcia-Sanchez, L.; Mourlon, C.; Nicoulaud, V.; Sy, M.; Gonze, M.A.

2015-01-01

SYMBIOSE is a modelling platform that accounts for variability and uncertainty in radiological impact assessments, when simulating the environmental fate of radionuclides and assessing doses to human populations. The default database of SYMBIOSE is partly based on parameter values that are summarized within International Atomic Energy Agency (IAEA) documents. To characterize uncertainty on the transfer parameters, 331 Probability Distribution Functions (PDFs) were defined from the summary statistics provided within the IAEA documents (i.e. sample size, minimal and maximum values, arithmetic and geometric means, standard and geometric standard deviations) and are made available as spreadsheet files. The methods used to derive the PDFs without complete data sets, but merely the summary statistics, are presented. Then, a simple case-study illustrates the use of the database in a second-order Monte Carlo calculation, separating parametric uncertainty and inter-individual variability. - Highlights: • Parametric uncertainty in radioecology was derived from IAEA documents. • 331 Probability Distribution Functions were defined for transfer parameters. • Parametric uncertainty and inter-individual variability were propagated

Development and assessment of best estimate integrated safety analysis code

International Nuclear Information System (INIS)

Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu

2007-03-01

Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published

Development and assessment of best estimate integrated safety analysis code

Energy Technology Data Exchange (ETDEWEB)

Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

2007-03-15

Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Guidance for treatment of variability and uncertainty in ecological risk assessments of contaminated sites

International Nuclear Information System (INIS)

1998-06-01

Uncertainty is a seemingly simple concept that has caused great confusion and conflict in the field of risk assessment. This report offers guidance for the analysis and presentation of variability and uncertainty in ecological risk assessments, an important issue in the remedial investigation and feasibility study processes. This report discusses concepts of probability in terms of variance and uncertainty, describes how these concepts differ in ecological risk assessment from human health risk assessment, and describes probabilistic aspects of specific ecological risk assessment techniques. The report ends with 17 points to consider in performing an uncertainty analysis for an ecological risk assessment of a contaminated site

Sensitivity-Uncertainty Techniques for Nuclear Criticality Safety

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-07

The sensitivity and uncertainty analysis course will introduce students to k_eff sensitivity data, cross-section uncertainty data, how k_eff sensitivity data and k_eff uncertainty data are generated and how they can be used. Discussion will include how sensitivity/uncertainty data can be used to select applicable critical experiments, to quantify a defensible margin to cover validation gaps and weaknesses, and in development of upper subcritical limits.

Exploring the uncertainties in cancer risk assessment using the integrated probabilistic risk assessment (IPRA) approach.

Science.gov (United States)

Slob, Wout; Bakker, Martine I; Biesebeek, Jan Dirk Te; Bokkers, Bas G H

2014-08-01

Current methods for cancer risk assessment result in single values, without any quantitative information on the uncertainties in these values. Therefore, single risk values could easily be overinterpreted. In this study, we discuss a full probabilistic cancer risk assessment approach in which all the generally recognized uncertainties in both exposure and hazard assessment are quantitatively characterized and probabilistically evaluated, resulting in a confidence interval for the final risk estimate. The methodology is applied to three example chemicals (aflatoxin, N-nitrosodimethylamine, and methyleugenol). These examples illustrate that the uncertainty in a cancer risk estimate may be huge, making single value estimates of cancer risk meaningless. Further, a risk based on linear extrapolation tends to be lower than the upper 95% confidence limit of a probabilistic risk estimate, and in that sense it is not conservative. Our conceptual analysis showed that there are two possible basic approaches for cancer risk assessment, depending on the interpretation of the dose-incidence data measured in animals. However, it remains unclear which of the two interpretations is the more adequate one, adding an additional uncertainty to the already huge confidence intervals for cancer risk estimates. © 2014 Society for Risk Analysis.

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

International Nuclear Information System (INIS)

Jones, G.

2013-01-01

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

Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty; Qualificacao e aplicacao de codigo de acidentes de reatores nucleares com capacidade interna de avaliacao de incerteza

Energy Technology Data Exchange (ETDEWEB)

Borges, Ronaldo Celem

2001-10-15

This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

Uncertainties on hydrocarbon exploration assessments in both the absence and presence of optioning

International Nuclear Information System (INIS)

Lerche, I.

1998-01-01

For hydrocarbon exploration opportunities a decision tree evaluation including variance in expected value leads to an extra uncertainty on the quality and worth of expected values as a decision device, due to both intrinsic uncertainties in success probability, assessed gains and assessed costs, and to the fact that the expected value is not one of the realizable outcomes. This paper shows how these uncertainty factors can be properly taken into account to provide a revised assessment of worth. In addition, a similar sense of logic prevails when options are considered for an opportunity. The uncertainty and success probability for an optional opportunity are also assessed in terms of the volatility of the maximum option worth. (author)

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

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

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

Science.gov (United States)

Herrero, J. J.; Ochoa, R.; Martínez, J. S.; Díez, C. J.; García-Herranz, N.; Cabellos, O.

2014-04-01

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Assessing flood forecast uncertainty with fuzzy arithmetic

Directory of Open Access Journals (Sweden)

de Bruyn Bertrand

2016-01-01

Full Text Available Providing forecasts for flow rates and water levels during floods have to be associated with uncertainty estimates. The forecast sources of uncertainty are plural. For hydrological forecasts (rainfall-runoff performed using a deterministic hydrological model with basic physics, two main sources can be identified. The first obvious source is the forcing data: rainfall forecast data are supplied in real time by meteorological forecasting services to the Flood Forecasting Service within a range between a lowest and a highest predicted discharge. These two values define an uncertainty interval for the rainfall variable provided on a given watershed. The second source of uncertainty is related to the complexity of the modeled system (the catchment impacted by the hydro-meteorological phenomenon, the number of variables that may describe the problem and their spatial and time variability. The model simplifies the system by reducing the number of variables to a few parameters. Thus it contains an intrinsic uncertainty. This model uncertainty is assessed by comparing simulated and observed rates for a large number of hydro-meteorological events. We propose a method based on fuzzy arithmetic to estimate the possible range of flow rates (and levels of water making a forecast based on possible rainfalls provided by forcing and uncertainty model. The model uncertainty is here expressed as a range of possible values. Both rainfall and model uncertainties are combined with fuzzy arithmetic. This method allows to evaluate the prediction uncertainty range. The Flood Forecasting Service of Oise and Aisne rivers, in particular, monitors the upstream watershed of the Oise at Hirson. This watershed’s area is 310 km2. Its response time is about 10 hours. Several hydrological models are calibrated for flood forecasting in this watershed and use the rainfall forecast. This method presents the advantage to be easily implemented. Moreover, it permits to be carried out

Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

Directory of Open Access Journals (Sweden)

Enrico Zio

2008-01-01

Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

Use of reliability engineering tools in safety and risk assessment of nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias, E-mail: amandaraso@hotmail.com, E-mail: vasconv@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: soaresw@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Serviço de Tecnologia de Reatores

2017-07-01

Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

Use of reliability engineering tools in safety and risk assessment of nuclear facilities

International Nuclear Information System (INIS)

Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias

2017-01-01

Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

LISA package user guide. Part II: LISA (Long Term Isolation Safety Assessment) program description and user guide

International Nuclear Information System (INIS)

Prado, P.; Saltelli, A.; Homma, T.

1992-01-01

This manual is subdivided into three parts. In this second part, this document describes the LISA (Long term Isolation Safety Assessment) Code and its submodels. LISA is a tool for analysis of the safety of an underground disposal of nuclear waste. It has the capability to handle nuclide chain of arbitrary length and to evaluate the migration of nuclide through a geosphere medium composed of an arbitrary number of segments. LISA makes use of Monte Carlo methodology to evaluate the uncertainty in the quantity being assessed (eg dose) arising from the uncertainty in the model input parameters. In the present version LISA is equipped with a very simple source term submodel, a relatively complex geosphere and a simplified biosphere. The code is closely associated with its statistical pre-processor code (PREP), which generates the input Monte Carlo sample from the assigned parameter probability density functions and with its post-processor code (SPOP) which provides useful statistics on the output sample (uncertainty and sensitivity analysis). This report describes the general structure of LISA, its subroutines and submodels, the code input ant output files. It is intended to provide the user with enough information to know and run the code as well as the capacity to incorporate different submodels. 15 refs., 6 figs

Development and comparison of Bayesian modularization method in uncertainty assessment of hydrological models

Science.gov (United States)

Li, L.; Xu, C.-Y.; Engeland, K.

2012-04-01

With respect to model calibration, parameter estimation and analysis of uncertainty sources, different approaches have been used in hydrological models. Bayesian method is one of the most widely used methods for uncertainty assessment of hydrological models, which incorporates different sources of information into a single analysis through Bayesian theorem. However, none of these applications can well treat the uncertainty in extreme flows of hydrological models' simulations. This study proposes a Bayesian modularization method approach in uncertainty assessment of conceptual hydrological models by considering the extreme flows. It includes a comprehensive comparison and evaluation of uncertainty assessments by a new Bayesian modularization method approach and traditional Bayesian models using the Metropolis Hasting (MH) algorithm with the daily hydrological model WASMOD. Three likelihood functions are used in combination with traditional Bayesian: the AR (1) plus Normal and time period independent model (Model 1), the AR (1) plus Normal and time period dependent model (Model 2) and the AR (1) plus multi-normal model (Model 3). The results reveal that (1) the simulations derived from Bayesian modularization method are more accurate with the highest Nash-Sutcliffe efficiency value, and (2) the Bayesian modularization method performs best in uncertainty estimates of entire flows and in terms of the application and computational efficiency. The study thus introduces a new approach for reducing the extreme flow's effect on the discharge uncertainty assessment of hydrological models via Bayesian. Keywords: extreme flow, uncertainty assessment, Bayesian modularization, hydrological model, WASMOD

Probability and Confidence Trade-space (PACT) Evaluation: Accounting for Uncertainty in Sparing Assessments

Science.gov (United States)

Anderson, Leif; Box, Neil; Carter, Katrina; DiFilippo, Denise; Harrington, Sean; Jackson, David; Lutomski, Michael

2012-01-01

There are two general shortcomings to the current annual sparing assessment: 1. The vehicle functions are currently assessed according to confidence targets, which can be misleading- overly conservative or optimistic. 2. The current confidence levels are arbitrarily determined and do not account for epistemic uncertainty (lack of knowledge) in the ORU failure rate. There are two major categories of uncertainty that impact Sparing Assessment: (a) Aleatory Uncertainty: Natural variability in distribution of actual failures around an Mean Time Between Failure (MTBF) (b) Epistemic Uncertainty : Lack of knowledge about the true value of an Orbital Replacement Unit's (ORU) MTBF We propose an approach to revise confidence targets and account for both categories of uncertainty, an approach we call Probability and Confidence Trade-space (PACT) evaluation.

Trends in development of probability assessment of nuclear power plant safety

International Nuclear Information System (INIS)

Dach, K.

1989-01-01

A complete study of probability safety assessment (PSA) of nuclear power plants is a multidisciplinary endeavor, requiring a qualified decision-making team composed of experienced professionals in individual disciplines and requiring good coordination of effort. The main concerns for the execution of a PSA study and related tasks are schematically presented. Also shown is a summary of the main steps for a PSA study at all three levels, with the incorporation of analysis of external events and the reliability of humans, including the necessary uncertainty analyses. 1 ref., 2 figs., 3 tabs

Analytical Propagation of Uncertainty in Life Cycle Assessment Using Matrix Formulation

DEFF Research Database (Denmark)

Imbeault-Tétreault, Hugues; Jolliet, Olivier; Deschênes, Louise

2013-01-01

with Monte Carlo results. The sensitivity and contribution of input parameters to output uncertainty were also analytically calculated. This article outlines an uncertainty analysis of the comparison between two case study scenarios. We conclude that the analytical method provides a good approximation...... on uncertainty calculation. This article shows the importance of the analytical method in uncertainty calculation, which could lead to a more complete uncertainty analysis in LCA practice....... uncertainty assessment is not a regular step in LCA. An analytical approach based on Taylor series expansion constitutes an effective means to overcome the drawbacks of the Monte Carlo method. This project aimed to test the approach on a real case study, and the resulting analytical uncertainty was compared...

Communicating uncertainty: lessons learned and suggestions for climate change assessment

International Nuclear Information System (INIS)

Patt, A.; Dessai, S.

2005-01-01

Assessments of climate change face the task of making information about uncertainty accessible and useful to decision-makers. The literature in behavior economics provides many examples of how people make decisions under conditions of uncertainty relying on inappropriate heuristics, leading to inconsistent and counterproductive choices. Modern risk communication practices recommend a number of methods to overcome these hurdles, which have been recommended for the Intergovernmental Panel on Climate Change (IPCC) assessment reports. This paper evaluates the success of the most recent IPCC approach to uncertainty communication, based on a controlled survey of climate change experts. Evaluating the results from the survey, and from a similar survey recently conducted among university students, the paper suggests that the most recent IPCC approach leaves open the possibility for biased and inconsistent responses to the information. The paper concludes by suggesting ways to improve the approach for future IPCC assessment reports. (authors)

Uncertainty in Impact Assessment – EIA in Denmark

DEFF Research Database (Denmark)

Larsen, Sanne Vammen

as problematic, as this is important information for decision makers and public actors. Taking point of departure in these issues, this paper seeks to add to the discussions by presenting the results of a study on the handling of uncertainty in Environmental Impact Assessment (EIA) reports in Denmark. The study...... is based on analysis of 100 EIA reports. The results will shed light on the extent to which uncertainties is addressed in EIA in Denmark and discuss how the practice can be categorised....

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.

Needs of the CSAU uncertainty method

International Nuclear Information System (INIS)

Prosek, A.; Mavko, B.

2000-01-01

The use of best estimate codes for safety analysis requires quantification of the uncertainties. These uncertainties are inherently linked to the chosen safety analysis methodology. Worldwide, various methods were proposed for this quantification. The purpose of this paper was to identify the needs of the Code Scaling, Applicability, and Uncertainty (CSAU) methodology and then to answer the needs. The specific procedural steps were combined from other methods for uncertainty evaluation and new tools and procedures were proposed. The uncertainty analysis approach and tools were then utilized for confirmatory study. The uncertainty was quantified for the RELAP5/MOD3.2 thermalhydraulic computer code. The results of the adapted CSAU approach to the small-break loss-of-coolant accident (SB LOCA) show that the adapted CSAU can be used for any thermal-hydraulic safety analysis with uncertainty evaluation. However, it was indicated that there are still some limitations in the CSAU approach that need to be resolved. (author)

Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

Directory of Open Access Journals (Sweden)

Franco Pierro

2009-01-01

Full Text Available The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks' formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

OSART Independent Safety Culture Assessment (ISCA) Guidelines

International Nuclear Information System (INIS)

2016-01-01

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

SFR inverse modelling Part 2. Uncertainty factors of predicted flow in deposition tunnels and uncertainty in distribution of flow paths from deposition tunnels

International Nuclear Information System (INIS)

Holmen, Johan

2007-10-01

The Swedish Nuclear Fuel and Waste Management Co (SKB) is operating the SFR repository for low- and intermediate-level nuclear waste. An update of the safety analysis of SFR was carried out by SKB as the SAFE project (Safety Assessment of Final Disposal of Operational Radioactive Waste). The aim of the project was to update the safety analysis and to produce a safety report. The safety report has been submitted to the Swedish authorities. This study is a continuation of the SAFE project, and concerns the hydrogeological modelling of the SFR repository, which was carried out as part of the SAFE project, it describes the uncertainty in the tunnel flow and distributions of flow paths from the storage tunnels. Uncertainty factors are produced for two different flow situations, corresponding to 2,000 AD (the sea covers the repository) and 4,000 AD (the sea has retreated form the repository area). Uncertainty factors are produced for the different deposition tunnels. The uncertainty factors are discussed in Chapter 2 and two lists (matrix) of uncertainty factors have been delivered as a part of this study. Flow paths are produced for two different flow situations, corresponding to 2,000 AD (the sea covers the repository) and 5,000 AD (the sea has retreated form the repository area). Flow paths from the different deposition tunnels have been simulated, considering the above discussed base case and the 60 realisation that passed all tests of this base case. The flow paths are presented and discussed in Chapter 3 and files presenting the results of the flow path analyses have been delivered as part of this study. The uncertainty factors (see Chapter 2) are not independent from the flow path data (see Chapter 3). When stochastic calculations are performed by use of a transport model and the data presented in this study is used as input to such calculations, the corresponding uncertainty factors and flow path data should be used. This study also includes a brief discussion of

Assessing student understanding of measurement and uncertainty

Science.gov (United States)

Jirungnimitsakul, S.; Wattanakasiwich, P.

2017-09-01

The objectives of this study were to develop and assess student understanding of measurement and uncertainty. A test has been adapted and translated from the Laboratory Data Analysis Instrument (LDAI) test, consists of 25 questions focused on three topics including measures of central tendency, experimental errors and uncertainties, and fitting regression lines. The test was evaluated its content validity by three physics experts in teaching physics laboratory. In the pilot study, Thai LDAI was administered to 93 freshmen enrolled in a fundamental physics laboratory course. The final draft of the test was administered to three groups—45 freshmen taking fundamental physics laboratory, 16 sophomores taking intermediated physics laboratory and 21 juniors taking advanced physics laboratory at Chiang Mai University. As results, we found that the freshmen had difficulties in experimental errors and uncertainties. Most students had problems with fitting regression lines. These results will be used to improve teaching and learning physics laboratory for physics students in the department.

Background and Qualification of Uncertainty Methods

International Nuclear Information System (INIS)

D'Auria, F.; Petruzzi, A.

2008-01-01

The evaluation of uncertainty constitutes the necessary supplement of Best Estimate calculations performed to understand accident scenarios in water cooled nuclear reactors. The needs come from the imperfection of computational tools on the one side and from the interest in using such tool to get more precise evaluation of safety margins. The paper reviews the salient features of two independent approaches for estimating uncertainties associated with predictions of complex system codes. Namely the propagation of code input error and the propagation of the calculation output error constitute the key-words for identifying the methods of current interest for industrial applications. Throughout the developed methods, uncertainty bands can be derived (both upper and lower) for any desired quantity of the transient of interest. For the second case, the uncertainty method is coupled with the thermal-hydraulic code to get the Code with capability of Internal Assessment of Uncertainty, whose features are discussed in more detail.

Metrics design for safety assessment

NARCIS (Netherlands)

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

2016-01-01

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

Status of uncertainty assessment in k0-NAA measurement. Anything still missing?

International Nuclear Information System (INIS)

Borut Smodis; Tinkara Bucar

2014-01-01

Several approaches to quantifying measurement uncertainty in k 0 -based neutron activation analysis (k 0 -NAA) are reviewed, comprising the original approach, the spreadsheet approach, the dedicated computer program involving analytical calculations and the two k 0 -NAA programs available on the market. Two imperfectness in the dedicated programs are identified, their impact assessed and possible improvements presented for a concrete experimental situation. The status of uncertainty assessment in k 0 -NAA is discussed and steps for improvement are recommended. It is concluded that the present magnitude of measurement uncertainty should further be improved by making additional efforts in reducing uncertainties of the relevant nuclear constants used. (author)

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

Science.gov (United States)

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

2013-12-01

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

Uncertainty assessing of measure result of tungsten in U3O8 by ICP-AES

International Nuclear Information System (INIS)

Du Guirong; Nie Jie; Tang Lilei

2011-01-01

According as the determining method and the assessing criterion,the uncertainty assessing of measure result of tungsten in U 3 O 8 by ICP-AES is researched. With the assessment of each component in detail, the result shows that u rel (sc)> u rel (c)> u rel (F)> u rel (m) by uncertainty contribution. Other uncertainty is random, calculated by repetition. u rel (sc) is contributed to uncertainty mainly. So the general uncertainty is reduced with strict operation to reduce u rel (sc). (authors)

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

Assessing performance of flaw characterization methods through uncertainty propagation

Science.gov (United States)

Miorelli, R.; Le Bourdais, F.; Artusi, X.

2018-04-01

In this work, we assess the inversion performance in terms of crack characterization and localization based on synthetic signals associated to ultrasonic and eddy current physics. More precisely, two different standard iterative inversion algorithms are used to minimize the discrepancy between measurements (i.e., the tested data) and simulations. Furthermore, in order to speed up the computational time and get rid of the computational burden often associated to iterative inversion algorithms, we replace the standard forward solver by a suitable metamodel fit on a database built offline. In a second step, we assess the inversion performance by adding uncertainties on a subset of the database parameters and then, through the metamodel, we propagate these uncertainties within the inversion procedure. The fast propagation of uncertainties enables efficiently evaluating the impact due to the lack of knowledge on some parameters employed to describe the inspection scenarios, which is a situation commonly encountered in the industrial NDE context.

Conceptual and computational basis for the quantification of margins and uncertainty

International Nuclear Information System (INIS)

Helton, Jon Craig

2009-01-01

In 2001, the National Nuclear Security Administration of the U.S. Department of Energy in conjunction with the national security laboratories (i.e, Los Alamos National Laboratory, Lawrence Livermore National Laboratory and Sandia National Laboratories) initiated development of a process designated Quantification of Margins and Uncertainty (QMU) for the use of risk assessment methodologies in the certification of the reliability and safety of the nation's nuclear weapons stockpile. This presentation discusses and illustrates the conceptual and computational basis of QMU in analyses that use computational models to predict the behavior of complex systems. Topics considered include (1) the role of aleatory and epistemic uncertainty in QMU, (2) the representation of uncertainty with probability, (3) the probabilistic representation of uncertainty in QMU analyses involving only epistemic uncertainty, (4) the probabilistic representation of uncertainty in QMU analyses involving aleatory and epistemic uncertainty, (5) procedures for sampling-based uncertainty and sensitivity analysis, (6) the representation of uncertainty with alternatives to probability such as interval analysis, possibility theory and evidence theory, (7) the representation of uncertainty with alternatives to probability in QMU analyses involving only epistemic uncertainty, and (8) the representation of uncertainty with alternatives to probability in QMU analyses involving aleatory and epistemic uncertainty. Concepts and computational procedures are illustrated with both notional examples and examples from reactor safety and radioactive waste disposal.

Safety assessment principles for nuclear plants

International Nuclear Information System (INIS)

1992-01-01

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

Development of safety related technology and infrastructure for safety assessment

International Nuclear Information System (INIS)

Venkat Raj, V.

1997-01-01

Development and optimum utilisation of any technology calls for the building up of the necessary infrastructure and backup facilities. This is particularly true for a developing country like India and more so for an advanced technology like nuclear technology. Right from the inception of its nuclear power programme, the Indian approach has been to develop adequate infrastructure in various areas such as design, construction, manufacture, installation, commissioning and safety assessment of nuclear plants. This paper deals with the development of safety related technology and the relevant infrastructure for safety assessment. A number of computer codes for safety assessment have been developed or adapted in the areas of thermal hydraulics, structural dynamics etc. These codes have undergone extensive validation through data generated in the experimental facilities set up in India as well as participation in international standard problem exercises. Side by side with the development of the tools for safety assessment, the development of safety related technology was also given equal importance. Many of the technologies required for the inspection, ageing assessment and estimation of the residual life of various components and equipment, particularly those having a bearing on safety, were developed. This paper highlights, briefly, the work carried out in some of the areas mentioned above. (author)

Thinking of the safety assessment of HLW disposal

International Nuclear Information System (INIS)

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

2014-01-01

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

Uncertainty Assessments in Fast Neutron Activation Analysis

International Nuclear Information System (INIS)

W. D. James; R. Zeisler

2000-01-01

Fast neutron activation analysis (FNAA) carried out with the use of small accelerator-based neutron generators is routinely used for major/minor element determinations in industry, mineral and petroleum exploration, and to some extent in research. While the method shares many of the operational procedures and therefore errors inherent to conventional thermal neutron activation analysis, its unique implementation gives rise to additional specific concerns that can result in errors or increased uncertainties of measured quantities. The authors were involved in a recent effort to evaluate irreversible incorporation of oxygen into a standard reference material (SRM) by direct measurement of oxygen by FNAA. That project required determination of oxygen in bottles of the SRM stored in varying environmental conditions and a comparison of the results. We recognized the need to accurately describe the total uncertainty of the measurements to accurately characterize any differences in the resulting average concentrations. It is our intent here to discuss the breadth of potential parameters that have the potential to contribute to the random and nonrandom errors of the method and provide estimates of the magnitude of uncertainty introduced. In addition, we will discuss the steps taken in this recent FNAA project to control quality, assess the uncertainty of the measurements, and evaluate results based on the statistical reproducibility

ALARP considerations in criticality safety assessments

International Nuclear Information System (INIS)

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

2003-01-01

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

Safety culture assessment developed by JANTI

International Nuclear Information System (INIS)

Hamada, Jun

2009-01-01

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

Environmental impact and risk assessments and key factors contributing to the overall uncertainties

International Nuclear Information System (INIS)

Salbu, Brit

2016-01-01

There is a significant number of nuclear and radiological sources that have contributed, are still contributing, or have the potential to contribute to radioactive contamination of the environment in the future. To protect the environment from radioactive contamination, impact and risk assessments are performed prior to or during a release event, short or long term after deposition or prior and after implementation of countermeasures. When environmental impact and risks are assessed, however, a series of factors will contribute to the overall uncertainties. To provide environmental impact and risk assessments, information on processes, kinetics and a series of input variables is needed. Adding problems such as variability, questionable assumptions, gaps in knowledge, extrapolations and poor conceptual model structures, a series of factors are contributing to large and often unacceptable uncertainties in impact and risk assessments. Information on the source term and the release scenario is an essential starting point in impact and risk models; the source determines activity concentrations and atom ratios of radionuclides released, while the release scenario determine the physico-chemical forms of released radionuclides such as particle size distribution, structure and density. Releases will most often contain other contaminants such as metals, and due to interactions, contaminated sites should be assessed as a multiple stressor scenario. Following deposition, a series of stressors, interactions and processes will influence the ecosystem transfer of radionuclide species and thereby influence biological uptake (toxicokinetics) and responses (toxicodynamics) in exposed organisms. Due to the variety of biological species, extrapolation is frequently needed to fill gaps in knowledge e.g., from effects to no effects, from effects in one organism to others, from one stressor to mixtures. Most toxtests are, however, performed as short term exposure of adult organisms

Uncertainties in human health risk assessment of environmental contaminants: A review and perspective.

Science.gov (United States)

Dong, Zhaomin; Liu, Yanju; Duan, Luchun; Bekele, Dawit; Naidu, Ravi

2015-12-01

Addressing uncertainties in human health risk assessment is a critical issue when evaluating the effects of contaminants on public health. A range of uncertainties exist through the source-to-outcome continuum, including exposure assessment, hazard and risk characterisation. While various strategies have been applied to characterising uncertainty, classical approaches largely rely on how to maximise the available resources. Expert judgement, defaults and tools for characterising quantitative uncertainty attempt to fill the gap between data and regulation requirements. The experiences of researching 2,3,7,8-tetrachlorodibenzo-p-dioxin (TCDD) illustrated uncertainty sources and how to maximise available information to determine uncertainties, and thereby provide an 'adequate' protection to contaminant exposure. As regulatory requirements and recurring issues increase, the assessment of complex scenarios involving a large number of chemicals requires more sophisticated tools. Recent advances in exposure and toxicology science provide a large data set for environmental contaminants and public health. In particular, biomonitoring information, in vitro data streams and computational toxicology are the crucial factors in the NexGen risk assessment, as well as uncertainties minimisation. Although in this review we cannot yet predict how the exposure science and modern toxicology will develop in the long-term, current techniques from emerging science can be integrated to improve decision-making. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

2012-04-01

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

Development of probabilistic methods for safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Schott, H.; Berg, H.P.

1998-01-01

Since its introduction by the German Risk Study, Probabilistic Safety Assessment (PSA) has developed in Germany to a valuable tool in regulatory decision-making. Plant specific PSAs of Level 1+ are now conducted for all nuclear power plants in the frame of Periodic Safety Reviews. This paper is devoted to the description or key elements set out in the regulatory guidelines for PSA-Level 1+ and the corresponding technical documents and the further development of PSA methodology in the Federal Republic of Germany. In the course of the next years it is intended to make progress in the modeling of common cause failures, human reliability evaluation, reduction of uncertainties in PSA modeling techniques and data estimation, analysis of low power and shut down states as well as in reaching a mature methodology for inclusion of external events into the analysis. (author)

Perspectives on dosimetric uncertainties and radiological assessments of radioactive waste management

International Nuclear Information System (INIS)

Smith, G.M.; Pinedo, P.; Cancio, D.

1997-01-01

The purpose of this paper is to raise some issues concerning uncertainties in the estimation of doses of ionizing radiation arising from waste management practices and the contribution to those uncertainties arising from dosimetry modelling. The intentions are: (a) to provide perspective on the relative uncertainties in the different aspects of radiological assessments of waste management; (b) to give pointers as to where resources could best be targeted as regards reduction in overall uncertainties; and (c) to provide regulatory insight to decisions on low dose management as related to waste management practices. (author)

Sensitivity to Uncertainty in Asteroid Impact Risk Assessment

Science.gov (United States)

Mathias, D.; Wheeler, L.; Prabhu, D. K.; Aftosmis, M.; Dotson, J.; Robertson, D. K.

2015-12-01

The Engineering Risk Assessment (ERA) team at NASA Ames Research Center is developing a physics-based impact risk model for probabilistically assessing threats from potential asteroid impacts on Earth. The model integrates probabilistic sampling of asteroid parameter ranges with physics-based analyses of entry, breakup, and impact to estimate damage areas and casualties from various impact scenarios. Assessing these threats is a highly coupled, dynamic problem involving significant uncertainties in the range of expected asteroid characteristics, how those characteristics may affect the level of damage, and the fidelity of various modeling approaches and assumptions. The presented model is used to explore the sensitivity of impact risk estimates to these uncertainties in order to gain insight into what additional data or modeling refinements are most important for producing effective, meaningful risk assessments. In the extreme cases of very small or very large impacts, the results are generally insensitive to many of the characterization and modeling assumptions. However, the nature of the sensitivity can change across moderate-sized impacts. Results will focus on the value of additional information in this critical, mid-size range, and how this additional data can support more robust mitigation decisions.

Uncertainty analysis in the applications of nuclear probabilistic risk assessment

International Nuclear Information System (INIS)

Le Duy, T.D.

2011-01-01

The aim of this thesis is to propose an approach to model parameter and model uncertainties affecting the results of risk indicators used in the applications of nuclear Probabilistic Risk assessment (PRA). After studying the limitations of the traditional probabilistic approach to represent uncertainty in PRA model, a new approach based on the Dempster-Shafer theory has been proposed. The uncertainty analysis process of the proposed approach consists in five main steps. The first step aims to model input parameter uncertainties by belief and plausibility functions according to the data PRA model. The second step involves the propagation of parameter uncertainties through the risk model to lay out the uncertainties associated with output risk indicators. The model uncertainty is then taken into account in the third step by considering possible alternative risk models. The fourth step is intended firstly to provide decision makers with information needed for decision making under uncertainty (parametric and model) and secondly to identify the input parameters that have significant uncertainty contributions on the result. The final step allows the process to be continued in loop by studying the updating of beliefs functions given new data. The proposed methodology was implemented on a real but simplified application of PRA model. (author)

LOFT uncertainty-analysis methodology

International Nuclear Information System (INIS)

Lassahn, G.D.

1983-01-01

The methodology used for uncertainty analyses of measurements in the Loss-of-Fluid Test (LOFT) nuclear-reactor-safety research program is described and compared with other methodologies established for performing uncertainty analyses

LOFT uncertainty-analysis methodology

International Nuclear Information System (INIS)

Lassahn, G.D.

1983-01-01

The methodology used for uncertainty analyses of measurements in the Loss-of-Fluid Test (LOFT) nuclear reactor safety research program is described and compared with other methodologies established for performing uncertainty analyses

Probabilistic safety assessment of Narora Atomic Power Project

International Nuclear Information System (INIS)

Babar, A.K.; Saraf, R.K.; Kakodkar, A.; Sanyasi Rao, V.V.S.

1989-01-01

Various safety studies on Pressurised Water and Boiling Water reactors have been conducted. However, a detailed report on probabilistic safety assessment (PSA) of PHWRs is not available. PSA level I results of the standardised 235 MWe PHWR under construction at Narora are presented herein. Fault Tree analysis of various initiating events (IEs), safety systems has been completed. Event Tree analysis has been performed for all the dominating IEs to identify the accident sequences and a list of the dominating accident sequences is included. Analysis has been carried out using Monte Carlo simulation to propagate the uncertanities in failure rate data. Further uncertainty analysis is extended to obtain distributions for the accident sequences and core damage frequency. Some noteworthy results of the study apart from the various design modifications incorporated during the design phase are: (i) The accident sequences resulting from station blackout are dominant contributors to the core damage frequency. (ii) Class-IV transients, small break LOCA are significant IEs. Main steam line break is likely to induce steam generator tube ruptures. (iii) Moderator circulation, fire fighting system, secondary steam relief are relatively important in core damage frequency reductions. (iv) Under accidental situations human errors are likely to be asociated with valving in shutdown cooling and fire fighting systems. (author). 14 tabs., 14 figs., 15 refs

Uncertainty studies and risk assessment for CO2 storage in geological formations

International Nuclear Information System (INIS)

Walter, Lena Sophie

2013-01-01

Carbon capture and storage (CCS) in deep geological formations is one possible option to mitigate the greenhouse gas effect by reducing CO 2 emissions into the atmosphere. The assessment of the risks related to CO 2 storage is an important task. Events such as CO 2 leakage and brine displacement could result in hazards for human health and the environment. In this thesis, a systematic and comprehensive risk assessment concept is presented to investigate various levels of uncertainties and to assess risks using numerical simulations. Depending on the risk and the processes, which should be assessed, very complex models, large model domains, large time scales, and many simulations runs for estimating probabilities are required. To reduce the resulting high computational costs, a model reduction technique (the arbitrary polynomial chaos expansion) and a method for model coupling in space are applied. The different levels of uncertainties are: statistical uncertainty in parameter distributions, scenario uncertainty, e.g. different geological features, and recognized ignorance due to assumptions in the conceptual model set-up. Recognized ignorance and scenario uncertainty are investigated by simulating well defined model set-ups and scenarios. According to damage values, which are defined as a model output, the set-ups and scenarios can be compared and ranked. For statistical uncertainty probabilities can be determined by running Monte Carlo simulations with the reduced model. The results are presented in various ways: e.g., mean damage, probability density function, cumulative distribution function, or an overall risk value by multiplying the damage with the probability. If the model output (damage) cannot be compared to provided criteria (e.g. water quality criteria), analytical approximations are presented to translate the damage into comparable values. The overall concept is applied for the risks related to brine displacement and infiltration into drinking water

Quantified Uncertainties in Comparative Life Cycle Assessment : What Can Be Concluded?

NARCIS (Netherlands)

Mendoza Beltran, Angelica; Prado, Valentina; Font Vivanco, David; Henriksson, Patrik J.G.; Guinée, Jeroen B.; Heijungs, Reinout

2018-01-01

Interpretation of comparative Life Cycle Assessment (LCA) results can be challenging in the presence of uncertainty. To aid in interpreting such results under the goal of any comparative LCA, we aim to provide guidance to practitioners by gaining insights into uncertainty-statistics methods (USMs).

The Safety Case and Safety Assessment for the Disposal of Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-09-15

This Safety Guide provides guidance and recommendations on meeting the safety requirements in respect of the safety case and supporting safety assessment for the disposal of radioactive waste. The safety case and supporting safety assessment provide the basis for demonstration of safety and for licensing of radioactive waste disposal facilities and assist and guide decisions on siting, design and operations. The safety case is also the main basis on which dialogue with interested parties is conducted and on which confidence in the safety of the disposal facility is developed. This Safety Guide is relevant for operating organizations preparing the safety case as well as for the regulatory body responsible for developing the regulations and regulatory guidance that determine the basis and scope of the safety case. Contents: 1. Introduction; 2. Demonstrating the safety of radioactive waste disposal; 3. Safety principles and safety requirements; 4. The safety case for disposal of radioactive waste; 5. Radiological impact assessment for the period after closure; 6. Specific issues; 7. Documentation and use of the safety case; 8. Regulatory review process.

Risk Assessment and Decision-Making under Uncertainty in Tunnel and Underground Engineering

Directory of Open Access Journals (Sweden)

Yuanpu Xia

2017-10-01

Full Text Available The impact of uncertainty on risk assessment and decision-making is increasingly being prioritized, especially for large geotechnical projects such as tunnels, where uncertainty is often the main source of risk. Epistemic uncertainty, which can be reduced, is the focus of attention. In this study, the existing entropy-risk decision model is first discussed and analyzed, and its deficiencies are improved upon and overcome. Then, this study addresses the fact that existing studies only consider parameter uncertainty and ignore the influence of the model uncertainty. Here, focus is on the issue of model uncertainty and differences in risk consciousness with different decision-makers. The utility theory is introduced in the model. Finally, a risk decision model is proposed based on the sensitivity analysis and the tolerance cost, which can improve decision-making efficiency. This research can provide guidance or reference for the evaluation and decision-making of complex systems engineering problems, and indicate a direction for further research of risk assessment and decision-making issues.

New challenges on uncertainty propagation assessment of flood risk analysis

Science.gov (United States)

Martins, Luciano; Aroca-Jiménez, Estefanía; Bodoque, José M.; Díez-Herrero, Andrés

2016-04-01

Natural hazards, such as floods, cause considerable damage to the human life, material and functional assets every year and around the World. Risk assessment procedures has associated a set of uncertainties, mainly of two types: natural, derived from stochastic character inherent in the flood process dynamics; and epistemic, that are associated with lack of knowledge or the bad procedures employed in the study of these processes. There are abundant scientific and technical literature on uncertainties estimation in each step of flood risk analysis (e.g. rainfall estimates, hydraulic modelling variables); but very few experience on the propagation of the uncertainties along the flood risk assessment. Therefore, epistemic uncertainties are the main goal of this work, in particular,understand the extension of the propagation of uncertainties throughout the process, starting with inundability studies until risk analysis, and how far does vary a proper analysis of the risk of flooding. These methodologies, such as Polynomial Chaos Theory (PCT), Method of Moments or Monte Carlo, are used to evaluate different sources of error, such as data records (precipitation gauges, flow gauges...), hydrologic and hydraulic modelling (inundation estimation), socio-demographic data (damage estimation) to evaluate the uncertainties propagation (UP) considered in design flood risk estimation both, in numerical and cartographic expression. In order to consider the total uncertainty and understand what factors are contributed most to the final uncertainty, we used the method of Polynomial Chaos Theory (PCT). It represents an interesting way to handle to inclusion of uncertainty in the modelling and simulation process. PCT allows for the development of a probabilistic model of the system in a deterministic setting. This is done by using random variables and polynomials to handle the effects of uncertainty. Method application results have a better robustness than traditional analysis

Can Bayesian Belief Networks help tackling conceptual model uncertainties in contaminated site risk assessment?

DEFF Research Database (Denmark)

Troldborg, Mads; Thomsen, Nanna Isbak; McKnight, Ursula S.

different conceptual models may describe the same contaminated site equally well. In many cases, conceptual model uncertainty has been shown to be one of the dominant sources for uncertainty and is therefore essential to account for when quantifying uncertainties in risk assessments. We present here......A key component in risk assessment of contaminated sites is the formulation of a conceptual site model. The conceptual model is a simplified representation of reality and forms the basis for the mathematical modelling of contaminant fate and transport at the site. A conceptual model should...... a Bayesian Belief Network (BBN) approach for evaluating the uncertainty in risk assessment of groundwater contamination from contaminated sites. The approach accounts for conceptual model uncertainty by considering multiple conceptual models, each of which represents an alternative interpretation of the site...

Safety assessment for facilities and activities. General safety requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF 6 ; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2010-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation.? read more The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Regulatory review of safety cases and safety assessments - associated challenges

International Nuclear Information System (INIS)

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

2006-01-01

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

Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

International Nuclear Information System (INIS)

Boyack, B.; Duffey, R.; Wilson, G.; Griffith, P.; Lellouche, G.; Levy, S.; Rohatgi, U.; Wulff, W.; Zuber, N.

1989-12-01

The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

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

Safety assessments for potential exposures

International Nuclear Information System (INIS)

Dunn, D.I.

2012-04-01

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

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

International Nuclear Information System (INIS)

Song, Tae Young

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

Confronting Uncertainty in Life Cycle Assessment Used for Decision Support

DEFF Research Database (Denmark)

Herrmann, Ivan Tengbjerg; Hauschild, Michael Zwicky; Sohn, Michael D.

2014-01-01

the decision maker (DM) in making the best possible choice for the environment. At present, some DMs do not trust the LCA to be a reliable decisionsupport tool—often because DMs consider the uncertainty of an LCA to be too large. The standard evaluation of uncertainty in LCAs is an ex-post approach that can...... regarding which type of LCA study to employ for the decision context at hand. This taxonomy enables the derivation of an LCA classification matrix to clearly identify and communicate the type of a given LCA. By relating the LCA classification matrix to statistical principles, we can also rank the different......The aim of this article is to help confront uncertainty in life cycle assessments (LCAs) used for decision support. LCAs offer a quantitative approach to assess environmental effects of products, technologies, and services and are conducted by an LCA practitioner or analyst (AN) to support...

Performance assessment studies for the long-term safety evaluation of radioactive waste disposal facility

International Nuclear Information System (INIS)

Bujoreanu, D.; Olteanu, M.; Bujoreanu, L.

2008-01-01

Especially during the last ten years, a part of Romanian research program 'Management of Radioactive Waste and Spent Fuel' was focused mainly on applicative research for the design of near-surface disposal facility, which intends to accommodate the low and intermediate radioactive waste generated from Cernavoda NPP. In this frame, our contribution was at the acquisition of technical data for the characterization of the future disposal facility. In the present, the project of the disposal facility, located on the Saligny site, near Cernavoda NPP, must be licensed. As regards to the safe disposal, the location of final disposal, the Saligny site, has been characterized through the five geological formations which contain potential routes for transport of radionuclide released from disposal facility, in the receiving zones(potential receiving zones), into liquid and gaseous phases. The technical characteristics of the disposal facility were adapted at the Romanian disposal concept using the reference data from IAEA technical report (IAEA,1999). Input parameters which characterized from physical and chemical point of view the disposal system, were partially taken from literature. The performance assessment studies, which follows the preliminary design development phases and the selection, describes how the source term is affected by the infiltration of water through the disposal facility, degradation process of engineering barriers (reflected in the distribution coefficient values) and solubility limit. The studies regard the evaluation of the source term, sensitivity and uncertainty analysis provide the information on 'how' and 'why' were evaluated, following: (i) radiological safety assessment of near-surface disposal facility on Saligny site; (ii) complexity standard assessment of the Engineering Barriers Systems (EBS); (iii) identification of the elements which must be elaborated for the increase of the disposal safety and the necessity for new technical data for

On treatment of uncertainty in system planning

International Nuclear Information System (INIS)

Flage, R.; Aven, T.

2009-01-01

In system planning and operation considerable efforts and resources are spent to reduce uncertainties, as a part of project management, uncertainty management and safety management. The basic idea seems to be that uncertainties are purely negative and should be reduced. In this paper we challenge this way of thinking, using a common industry practice as an example. In accordance with this industry practice, three uncertainty interval categories are used: ±40% intervals for the feasibility phase, ±30% intervals for the concept development phase and ±20% intervals for the engineering phase. The problem is that such a regime could easily lead to a conservative management regime encouraging the use of existing methods and tools, as new activities and novel solutions and arrangements necessarily mean increased uncertainties. In the paper we suggest an alternative approach based on uncertainty and risk descriptions, but having no predefined uncertainty reduction structures. The approach makes use of risk assessments and economic optimisation tools such as the expected net present value, but acknowledges the need for broad risk management processes which extend beyond the analyses. Different concerns need to be balanced, including economic aspects, uncertainties and risk, and practicability

Uncertainty assessment in geodetic network adjustment by combining GUM and Monte-Carlo-simulations

Science.gov (United States)

Niemeier, Wolfgang; Tengen, Dieter

2017-06-01

In this article first ideas are presented to extend the classical concept of geodetic network adjustment by introducing a new method for uncertainty assessment as two-step analysis. In the first step the raw data and possible influencing factors are analyzed using uncertainty modeling according to GUM (Guidelines to the Expression of Uncertainty in Measurements). This approach is well established in metrology, but rarely adapted within Geodesy. The second step consists of Monte-Carlo-Simulations (MC-simulations) for the complete processing chain from raw input data and pre-processing to adjustment computations and quality assessment. To perform these simulations, possible realizations of raw data and the influencing factors are generated, using probability distributions for all variables and the established concept of pseudo-random number generators. Final result is a point cloud which represents the uncertainty of the estimated coordinates; a confidence region can be assigned to these point clouds, as well. This concept may replace the common concept of variance propagation and the quality assessment of adjustment parameters by using their covariance matrix. It allows a new way for uncertainty assessment in accordance with the GUM concept for uncertainty modelling and propagation. As practical example the local tie network in "Metsähovi Fundamental Station", Finland is used, where classical geodetic observations are combined with GNSS data.

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....

Modifications of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany based upon new version of Emergency Operating Procedures

International Nuclear Information System (INIS)

Aldorf, R.

1997-01-01

In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to reflect on Probabilistic Safety Assessment-1 basis on impact of Emergency Response Guidelines (as one particular event from the list of other modifications) on Plant Safety. Following highlights help to orient the reader in main general aspects, findings and issues of the work that currently continues on. Older results of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany have revealed that human behaviour during accident progression scenarios represent one of the most important aspects in plant safety. Current effort of Nuclear Power Plants Dukovany (Czech Republic) and Bohunice (Slovak Republic) is focussed on development of qualitatively new symptom-based Emergency Operating Procedures called Emergency Response Guidelines Supplier - Westinghouse Energy Systems Europe, Brussels works in cooperation with teams of specialist from both Nuclear Power Plants. In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to prove on Probabilistic Safety Assessment -1 basis an expected - positive impact of Emergency Response Guidelines on Plant Safety, Since this contract is currently still in progress, it is possible to release only preliminary conclusions and observations. Emergency Response Guidelines compare to original Emergency Operating Procedures substantially reduce uncertainty of general human behaviour during plant response to an accident process. It is possible to conclude that from the current scope Probabilistic Safety Assessment Dukovany point of view (until core damage), Emergency Response Guidelines represent adequately wide basis for mitigating any initiating event

Assessment of dose measurement uncertainty using RisøScan

DEFF Research Database (Denmark)

Helt-Hansen, J.; Miller, A.

2006-01-01

The dose measurement uncertainty of the dosimeter system RisoScan, office scanner and Riso B3 dosimeters has been assessed by comparison with spectrophotometer measurements of the same dosimeters. The reproducibility and the combined uncertainty were found to be approximately 2% and 4%, respectiv......%, respectively, at one standard deviation. The subroutine in RisoScan for electron energy measurement is shown to give results that are equivalent to the measurements with a scanning spectrophotometer. (c) 2006 Elsevier Ltd. All rights reserved....

Safety assessment for spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

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

Assessing measurement uncertainty in meteorology in urban environments

International Nuclear Information System (INIS)

Curci, S; Lavecchia, C; Frustaci, G; Pilati, S; Paganelli, C; Paolini, R

2017-01-01

Measurement uncertainty in meteorology has been addressed in a number of recent projects. In urban environments, uncertainty is also affected by local effects which are more difficult to deal with than for synoptic stations. In Italy, beginning in 2010, an urban meteorological network (Climate Network ® ) was designed, set up and managed at national level according to high metrological standards and homogeneity criteria to support energy applications. The availability of such a high-quality operative automatic weather station network represents an opportunity to investigate the effects of station siting and sensor exposure and to estimate the related measurement uncertainty. An extended metadata set was established for the stations in Milan, including siting and exposure details. Statistical analysis on an almost 3-year-long operational period assessed network homogeneity, quality and reliability. Deviations from reference mean values were then evaluated in selected low-gradient local weather situations in order to investigate siting and exposure effects. In this paper the methodology is depicted and preliminary results of its application to air temperature discussed; this allowed the setting of an upper limit of 1 °C for the added measurement uncertainty at the top of the urban canopy layer. (paper)

Assessing measurement uncertainty in meteorology in urban environments

Science.gov (United States)

Curci, S.; Lavecchia, C.; Frustaci, G.; Paolini, R.; Pilati, S.; Paganelli, C.

2017-10-01

Measurement uncertainty in meteorology has been addressed in a number of recent projects. In urban environments, uncertainty is also affected by local effects which are more difficult to deal with than for synoptic stations. In Italy, beginning in 2010, an urban meteorological network (Climate Network®) was designed, set up and managed at national level according to high metrological standards and homogeneity criteria to support energy applications. The availability of such a high-quality operative automatic weather station network represents an opportunity to investigate the effects of station siting and sensor exposure and to estimate the related measurement uncertainty. An extended metadata set was established for the stations in Milan, including siting and exposure details. Statistical analysis on an almost 3-year-long operational period assessed network homogeneity, quality and reliability. Deviations from reference mean values were then evaluated in selected low-gradient local weather situations in order to investigate siting and exposure effects. In this paper the methodology is depicted and preliminary results of its application to air temperature discussed; this allowed the setting of an upper limit of 1 °C for the added measurement uncertainty at the top of the urban canopy layer.

An uncertainty inventory demonstration - a primary step in uncertainty quantification

Energy Technology Data Exchange (ETDEWEB)

Langenbrunner, James R. [Los Alamos National Laboratory; Booker, Jane M [Los Alamos National Laboratory; Hemez, Francois M [Los Alamos National Laboratory; Salazar, Issac F [Los Alamos National Laboratory; Ross, Timothy J [UNM

2009-01-01

Tools, methods, and theories for assessing and quantifying uncertainties vary by application. Uncertainty quantification tasks have unique desiderata and circumstances. To realistically assess uncertainty requires the engineer/scientist to specify mathematical models, the physical phenomena of interest, and the theory or framework for assessments. For example, Probabilistic Risk Assessment (PRA) specifically identifies uncertainties using probability theory, and therefore, PRA's lack formal procedures for quantifying uncertainties that are not probabilistic. The Phenomena Identification and Ranking Technique (PIRT) proceeds by ranking phenomena using scoring criteria that results in linguistic descriptors, such as importance ranked with words, 'High/Medium/Low.' The use of words allows PIRT to be flexible, but the analysis may then be difficult to combine with other uncertainty theories. We propose that a necessary step for the development of a procedure or protocol for uncertainty quantification (UQ) is the application of an Uncertainty Inventory. An Uncertainty Inventory should be considered and performed in the earliest stages of UQ.

Probabilistic Radiological Performance Assessment Modeling and Uncertainty

Science.gov (United States)

Tauxe, J.

2004-12-01

A generic probabilistic radiological Performance Assessment (PA) model is presented. The model, built using the GoldSim systems simulation software platform, concerns contaminant transport and dose estimation in support of decision making with uncertainty. Both the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE) require assessments of potential future risk to human receptors of disposal of LLW. Commercially operated LLW disposal facilities are licensed by the NRC (or agreement states), and the DOE operates such facilities for disposal of DOE-generated LLW. The type of PA model presented is probabilistic in nature, and hence reflects the current state of knowledge about the site by using probability distributions to capture what is expected (central tendency or average) and the uncertainty (e.g., standard deviation) associated with input parameters, and propagating through the model to arrive at output distributions that reflect expected performance and the overall uncertainty in the system. Estimates of contaminant release rates, concentrations in environmental media, and resulting doses to human receptors well into the future are made by running the model in Monte Carlo fashion, with each realization representing a possible combination of input parameter values. Statistical summaries of the results can be compared to regulatory performance objectives, and decision makers are better informed of the inherently uncertain aspects of the model which supports their decision-making. While this information may make some regulators uncomfortable, they must realize that uncertainties which were hidden in a deterministic analysis are revealed in a probabilistic analysis, and the chance of making a correct decision is now known rather than hoped for. The model includes many typical features and processes that would be part of a PA, but is entirely fictitious. This does not represent any particular site and is meant to be a generic example. A

Consideration of uncertainties in CCDF risk curves in safety oriented decision making processes

International Nuclear Information System (INIS)

Stern, E.; Tadmor, J.

1988-01-01

In recent years, some of the results of Probabilistic Risk Assessment (i.e. the magnitudes of the various adverse health effects and other effects of potential accidents in nuclear power plants) have usually been presented in Complementary Cumulative Distribution Function curves, widely known as CCDF risk curves. CCDF curves are characteristic of probabilistic accident analyses and consequence calculations, although, in many cases, the codes producing the CCDF curves consist of a mixture of both probabilistic and deterministic calculations. One of the main difficulties in the process of PRA is the problem of uncertainties associated with the risk assessments. The uncertainties, as expressed in CCDF risk curves can be classified into two main categories: (a) uncertainties expressed by the CCDF risk curve itself due to its probabilistic nature and - (b) the uncertainty band of CCDF risk curves. The band consists of a ''family of CCDF curves'' which represents the risks (e.g. early/late fatalities) evaluated at various levels of confidence for a specific Plant-Site Combination (PSC) i.e. a certain nuclear power plant located at a certain site. The reasons why a family of curves rather than a single curve represents the risk of a certain PSC have been discussed. Generally, the uncertainty band of CCDF curves is limited by the 95% (''conservative'') and the 5% curves. In most cases the 50% (median, ''best estimate'') curve is also shown because scientists tend to believe that it represents the ''realistic'' (or real) risk of the plant

Ecosystem Services Mapping Uncertainty Assessment: A Case Study in the Fitzroy Basin Mining Region

Directory of Open Access Journals (Sweden)

Zhenyu Wang

2018-01-01

Full Text Available Ecosystem services mapping is becoming increasingly popular through the use of various readily available mapping tools, however, uncertainties in assessment outputs are commonly ignored. Uncertainties from different sources have the potential to lower the accuracy of mapping outputs and reduce their reliability for decision-making. Using a case study in an Australian mining region, this paper assessed the impact of uncertainties on the modelling of the hydrological ecosystem service, water provision. Three types of uncertainty were modelled using multiple uncertainty scenarios: (1 spatial data sources; (2 modelling scales (temporal and spatial and (3 parameterization and model selection. We found that the mapping scales can induce significant changes to the spatial pattern of outputs and annual totals of water provision. In addition, differences in parameterization using differing sources from the literature also led to obvious differences in base flow. However, the impact of each uncertainty associated with differences in spatial data sources were not so great. The results of this study demonstrate the importance of uncertainty assessment and highlight that any conclusions drawn from ecosystem services mapping, such as the impacts of mining, are likely to also be a property of the uncertainty in ecosystem services mapping methods.

Reliability assessment of complex electromechanical systems under epistemic uncertainty

International Nuclear Information System (INIS)

Mi, Jinhua; Li, Yan-Feng; Yang, Yuan-Jian; Peng, Weiwen; Huang, Hong-Zhong

2016-01-01

The appearance of macro-engineering and mega-project have led to the increasing complexity of modern electromechanical systems (EMSs). The complexity of the system structure and failure mechanism makes it more difficult for reliability assessment of these systems. Uncertainty, dynamic and nonlinearity characteristics always exist in engineering systems due to the complexity introduced by the changing environments, lack of data and random interference. This paper presents a comprehensive study on the reliability assessment of complex systems. In view of the dynamic characteristics within the system, it makes use of the advantages of the dynamic fault tree (DFT) for characterizing system behaviors. The lifetime of system units can be expressed as bounded closed intervals by incorporating field failures, test data and design expertize. Then the coefficient of variation (COV) method is employed to estimate the parameters of life distributions. An extended probability-box (P-Box) is proposed to convey the present of epistemic uncertainty induced by the incomplete information about the data. By mapping the DFT into an equivalent Bayesian network (BN), relevant reliability parameters and indexes have been calculated. Furthermore, the Monte Carlo (MC) simulation method is utilized to compute the DFT model with consideration of system replacement policy. The results show that this integrated approach is more flexible and effective for assessing the reliability of complex dynamic systems. - Highlights: • A comprehensive study on the reliability assessment of complex system is presented. • An extended probability-box is proposed to convey the present of epistemic uncertainty. • The dynamic fault tree model is built. • Bayesian network and Monte Carlo simulation methods are used. • The reliability assessment of a complex electromechanical system is performed.

Consideration of aging in probabilistic safety assessment

International Nuclear Information System (INIS)

Titina, B.; Cepin, M.

2007-01-01

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

A Methodology for Safety Culture Impact Assessment

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

A Methodology for Safety Culture Impact Assessment

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2014-01-01

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

Understanding uncertainty propagation in life cycle assessments of waste management systems

DEFF Research Database (Denmark)

Bisinella, Valentina; Conradsen, Knut; Christensen, Thomas Højlund

2015-01-01

Uncertainty analysis in Life Cycle Assessments (LCAs) of waste management systems often results obscure and complex, with key parameters rarely determined on a case-by-case basis. The paper shows an application of a simplified approach to uncertainty coupled with a Global Sensitivity Analysis (GSA......) perspective on three alternative waste management systems for Danish single-family household waste. The approach provides a fast and systematic method to select the most important parameters in the LCAs, understand their propagation and contribution to uncertainty....

Study on uncertainty evaluation system for the safety evaluation of interim spent fuel storage facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Hyeon; Shin, Myeong Won; Rhy, Seok Jin; Cho, Dong Keon; Park, Dong Hwan [Kyunghee Univ., Seoul (Korea, Republic of); Cheong, Beom Jin [Minstry of Science and Technology, Gwacheon (Korea, Republic of)

1998-03-15

The main objective os to develop a technical standards for the facility operation of the interm, spent fuel storage facility and to develop a draft for the technical criteria to be legislated. The another objective os to define a uncertainty evaluation system for burn up credit application in criticality analysis and to investigate an applicability of this topic for future regulatory activity. Investigate a status of art for the operational criteria of spent fuel interm wet storage. Collect relevant laws, decree, notices and standards related to the operation of storage facility and study on the legislation system. Develop a draft of technical standards and criteria to be legislated. Define an evaluation system for the uncertainty analysis and study on the status of art in the field of criticality safety analysis. Develop an uncertainty evaluation system in criticality analysis with burnup credit and investigate an applicability as well as its benefits of this policy.

A risk assessment methodology for incorporating uncertainties using fuzzy concepts

International Nuclear Information System (INIS)

Cho, Hyo-Nam; Choi, Hyun-Ho; Kim, Yoon-Bae

2002-01-01

This paper proposes a new methodology for incorporating uncertainties using fuzzy concepts into conventional risk assessment frameworks. This paper also introduces new forms of fuzzy membership curves, designed to consider the uncertainty range that represents the degree of uncertainties involved in both probabilistic parameter estimates and subjective judgments, since it is often difficult or even impossible to precisely estimate the occurrence rate of an event in terms of one single crisp probability. It is to be noted that simple linguistic variables such as 'High/Low' and 'Good/Bad' have the limitations in quantifying the various risks inherent in construction projects, but only represent subjective mental cognition adequately. Therefore, in this paper, the statements that include some quantification with giving specific value or scale, such as 'Close to any value' or 'Higher/Lower than analyzed value', are used in order to get over the limitations. It may be stated that the proposed methodology will be very useful for the systematic and rational risk assessment of construction projects

Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

International Nuclear Information System (INIS)

Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

2014-01-01

Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

Regulatory review of safety cases and safety assessments for near surface

International Nuclear Information System (INIS)

Nys, V.

2003-01-01

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

Long term safety requirements and safety indicators for the assessment of underground radioactive waste repositories

International Nuclear Information System (INIS)

Vovk, Ivan

1998-01-01

This presentation defines: waste disposal, safety issues, risk estimation; describes the integrated waste disposal process including quality assurance program. Related to actinides inventory it shows the main results of calculated activity obtained by deterministic estimation. It includes the Radioactive Waste Safety Standards and requirements; features related to site, design and waste package characteristics, as technical long term safety criteria for radioactive waste disposal facilities. Fundamental concern regarding the safety of radioactive waste disposal systems is their radiological impact on human beings and the environment. Safety requirements and criteria for judging the level of safety of such systems have been developed and there is a consensus among the international community on their basis within the well-established system of radiological protection. So far, however, little experience has been gained in applying long term safety criteria to actual disposal systems; consequently, there is an international debate on the most appropriate nature and form of the criteria to be used, taking into account the uncertainties involved. Emerging from the debate is the increasing conviction that the combined use of a variety of indicators would be advantageous in addressing the issue of reasonable assurance in the different time frames involved and in supporting the safety case for any particular repository concept. Indicators including risk, dose, radionuclide concentration, transit time, toxicity indices, fluxes at different points within the system, and barrier performance have all been identified as potentially relevant. Dose and risk are the indicators generally seen as most fundamental, as they seek directly to describe the radiological impact of a disposal system, and these are the ones that have been incorporated into most national standards to date. There are, however, certain problems in applying them. Application of a variety of different indicators

Development and comparison in uncertainty assessment based Bayesian modularization method in hydrological modeling

Science.gov (United States)

Li, Lu; Xu, Chong-Yu; Engeland, Kolbjørn

2013-04-01

SummaryWith respect to model calibration, parameter estimation and analysis of uncertainty sources, various regression and probabilistic approaches are used in hydrological modeling. A family of Bayesian methods, which incorporates different sources of information into a single analysis through Bayes' theorem, is widely used for uncertainty assessment. However, none of these approaches can well treat the impact of high flows in hydrological modeling. This study proposes a Bayesian modularization uncertainty assessment approach in which the highest streamflow observations are treated as suspect information that should not influence the inference of the main bulk of the model parameters. This study includes a comprehensive comparison and evaluation of uncertainty assessments by our new Bayesian modularization method and standard Bayesian methods using the Metropolis-Hastings (MH) algorithm with the daily hydrological model WASMOD. Three likelihood functions were used in combination with standard Bayesian method: the AR(1) plus Normal model independent of time (Model 1), the AR(1) plus Normal model dependent on time (Model 2) and the AR(1) plus Multi-normal model (Model 3). The results reveal that the Bayesian modularization method provides the most accurate streamflow estimates measured by the Nash-Sutcliffe efficiency and provide the best in uncertainty estimates for low, medium and entire flows compared to standard Bayesian methods. The study thus provides a new approach for reducing the impact of high flows on the discharge uncertainty assessment of hydrological models via Bayesian method.

HTGR reactor physics, thermal-hydraulics and depletion uncertainty analysis: a proposed IAEA coordinated research project

International Nuclear Information System (INIS)

Tyobeka, Bismark; Reitsma, Frederik; Ivanov, Kostadin

2011-01-01

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)

Safety Auditing and Assessments

Science.gov (United States)

Goodin, James Ronald (Ronnie)

2005-01-01

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

Uncertainty in hydraulic tests in fractured rock

International Nuclear Information System (INIS)

Ji, Sung-Hoon; Koh, Yong-Kwon

2014-01-01

Interpretation of hydraulic tests in fractured rock has uncertainty because of the different hydraulic properties of a fractured rock to a porous medium. In this study, we reviewed several interesting phenomena which show uncertainty in a hydraulic test at a fractured rock and discussed their origins and the how they should be considered during site characterisation. Our results show that the estimated hydraulic parameters of a fractured rock from a hydraulic test are associated with uncertainty due to the changed aperture and non-linear groundwater flow during the test. Although the magnitude of these two uncertainties is site-dependent, the results suggest that it is recommended to conduct a hydraulic test with a little disturbance from the natural groundwater flow to consider their uncertainty. Other effects reported from laboratory and numerical experiments such as the trapping zone effect (Boutt, 2006) and the slip condition effect (Lee, 2014) can also introduce uncertainty to a hydraulic test, which should be evaluated in a field test. It is necessary to consider the way how to evaluate the uncertainty in the hydraulic property during the site characterisation and how to apply it to the safety assessment of a subsurface repository. (authors)

The safety assessment of OPR-1000 nuclear power plant for station blackout accident applying the combined deterministic and probabilistic procedure

Energy Technology Data Exchange (ETDEWEB)

Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2014-08-15

Highlights: • The combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. • The safety assessment of OPR-1000 nuclear power plant for SBO accident is performed by applying the CDPP. • By estimating the offsite power restoration time appropriately, the SBO risk is reevaluated. • It is concluded that the CDPP is applicable to safety assessment of BDBAs without significant erosion of the safety margin. - Abstract: Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

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

International Nuclear Information System (INIS)

2001-01-01

results of SR 97 have been applied to formulate requirements and preferences regarding the host rock for a repository. In the authorities' opinion, SR 97 does not include a description of how this has been done. The coupling between safety assessment and site investigation, should be improved. A safety assessment of a repository for spent nuclear fuel will always contain uncertainties and deficiencies in the underlying data. Access to experts who can provide expert judgement is therefore vital. SKB should improve its procedures for obtaining expert judgement. To summarize, the authorities find that parts of the methodology in SR 97 must be further developed and detailed prior to the forthcoming stages of the site selection process. SKB's development of methods for safety assessment is a continuous process that should be conducted throughout all of the stages of the final disposal process. On the basis of the review of SR 97 and previous reviews of SKB's Research, Development and Demonstration programmes the authorities find that the KBS-3 method is an adequate basis for SKB's forthcoming site investigations and for the further development of the engineered barriers. In connection with future reviews of SKB's R, D and D programmes, the authorities intend to present additional views on the reporting that is necessary prior to the different stages of SKB's final disposal programme

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

Energy Technology Data Exchange (ETDEWEB)

Ghofrani, Javad

2016-05-26

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

International Nuclear Information System (INIS)

Ghofrani, Javad

2016-01-01

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Uncertainty studies and risk assessment for CO{sub 2} storage in geological formations

Energy Technology Data Exchange (ETDEWEB)

Walter, Lena Sophie

2013-07-01

Carbon capture and storage (CCS) in deep geological formations is one possible option to mitigate the greenhouse gas effect by reducing CO{sub 2} emissions into the atmosphere. The assessment of the risks related to CO{sub 2} storage is an important task. Events such as CO{sub 2} leakage and brine displacement could result in hazards for human health and the environment. In this thesis, a systematic and comprehensive risk assessment concept is presented to investigate various levels of uncertainties and to assess risks using numerical simulations. Depending on the risk and the processes, which should be assessed, very complex models, large model domains, large time scales, and many simulations runs for estimating probabilities are required. To reduce the resulting high computational costs, a model reduction technique (the arbitrary polynomial chaos expansion) and a method for model coupling in space are applied. The different levels of uncertainties are: statistical uncertainty in parameter distributions, scenario uncertainty, e.g. different geological features, and recognized ignorance due to assumptions in the conceptual model set-up. Recognized ignorance and scenario uncertainty are investigated by simulating well defined model set-ups and scenarios. According to damage values, which are defined as a model output, the set-ups and scenarios can be compared and ranked. For statistical uncertainty probabilities can be determined by running Monte Carlo simulations with the reduced model. The results are presented in various ways: e.g., mean damage, probability density function, cumulative distribution function, or an overall risk value by multiplying the damage with the probability. If the model output (damage) cannot be compared to provided criteria (e.g. water quality criteria), analytical approximations are presented to translate the damage into comparable values. The overall concept is applied for the risks related to brine displacement and infiltration into

Implications of Monte Carlo Statistical Errors in Criticality Safety Assessments

International Nuclear Information System (INIS)

Pevey, Ronald E.

2005-01-01

Most criticality safety calculations are performed using Monte Carlo techniques because of Monte Carlo's ability to handle complex three-dimensional geometries. For Monte Carlo calculations, the more histories sampled, the lower the standard deviation of the resulting estimates. The common intuition is, therefore, that the more histories, the better; as a result, analysts tend to run Monte Carlo analyses as long as possible (or at least to a minimum acceptable uncertainty). For Monte Carlo criticality safety analyses, however, the optimization situation is complicated by the fact that procedures usually require that an extra margin of safety be added because of the statistical uncertainty of the Monte Carlo calculations. This additional safety margin affects the impact of the choice of the calculational standard deviation, both on production and on safety. This paper shows that, under the assumptions of normally distributed benchmarking calculational errors and exact compliance with the upper subcritical limit (USL), the standard deviation that optimizes production is zero, but there is a non-zero value of the calculational standard deviation that minimizes the risk of inadvertently labeling a supercritical configuration as subcritical. Furthermore, this value is shown to be a simple function of the typical benchmarking step outcomes--the bias, the standard deviation of the bias, the upper subcritical limit, and the number of standard deviations added to calculated k-effectives before comparison to the USL

THE UNCERTAINTIES ON THE GIS BASED LAND SUITABILITY ASSESSMENT FOR URBAN AND RURAL PLANNING

Directory of Open Access Journals (Sweden)

H. Liu

2017-09-01

Full Text Available The majority of the research on the uncertainties of spatial data and spatial analysis focuses on some specific data feature or analysis tool. Few have accomplished the uncertainties of the whole process of an application like planning, making the research of uncertainties detached from practical applications. The paper discusses the uncertainties of the geographical information systems (GIS based land suitability assessment in planning on the basis of literature review. The uncertainties considered range from index system establishment to the classification of the final result. Methods to reduce the uncertainties arise from the discretization of continuous raster data and the index weight determination are summarized. The paper analyzes the merits and demerits of the “Nature Breaks” method which is broadly used by planners. It also explores the other factors which impact the accuracy of the final classification like the selection of class numbers, intervals and the autocorrelation of the spatial data. In the conclusion part, the paper indicates that the adoption of machine learning methods should be modified to integrate the complexity of land suitability assessment. The work contributes to the application of spatial data and spatial analysis uncertainty research on land suitability assessment, and promotes the scientific level of the later planning and decision-making.

Assessment of herbal medicinal products: Challenges, and opportunities to increase the knowledge base for safety assessment

International Nuclear Information System (INIS)

Jordan, Scott A.; Cunningham, David G.; Marles, Robin J.

2010-01-01

Although herbal medicinal products (HMP) have been perceived by the public as relatively low risk, there has been more recognition of the potential risks associated with this type of product as the use of HMPs increases. Potential harm can occur via inherent toxicity of herbs, as well as from contamination, adulteration, plant misidentification, and interactions with other herbal products or pharmaceutical drugs. Regulatory safety assessment for HMPs relies on both the assessment of cases of adverse reactions and the review of published toxicity information. However, the conduct of such an integrated investigation has many challenges in terms of the quantity and quality of information. Adverse reactions are under-reported, product quality may be less than ideal, herbs have a complex composition and there is lack of information on the toxicity of medicinal herbs or their constituents. Nevertheless, opportunities exist to capitalise on newer information to increase the current body of scientific evidence. Novel sources of information are reviewed, such as the use of poison control data to augment adverse reaction information from national pharmacovigilance databases, and the use of more recent toxicological assessment techniques such as predictive toxicology and omics. The integration of all available information can reduce the uncertainty in decision making with respect to herbal medicinal products. The example of Aristolochia and aristolochic acids is used to highlight the challenges related to safety assessment, and the opportunities that exist to more accurately elucidate the toxicity of herbal medicines.

A new computational method of a moment-independent uncertainty importance measure

International Nuclear Information System (INIS)

Liu Qiao; Homma, Toshimitsu

2009-01-01

For a risk assessment model, the uncertainty in input parameters is propagated through the model and leads to the uncertainty in the model output. The study of how the uncertainty in the output of a model can be apportioned to the uncertainty in the model inputs is the job of sensitivity analysis. Saltelli [Sensitivity analysis for importance assessment. Risk Analysis 2002;22(3):579-90] pointed out that a good sensitivity indicator should be global, quantitative and model free. Borgonovo [A new uncertainty importance measure. Reliability Engineering and System Safety 2007;92(6):771-84] further extended these three requirements by adding the fourth feature, moment-independence, and proposed a new sensitivity measure, δ i . It evaluates the influence of the input uncertainty on the entire output distribution without reference to any specific moment of the model output. In this paper, a new computational method of δ i is proposed. It is conceptually simple and easier to implement. The feasibility of this new method is proved by applying it to two examples.

Quantification of Wave Model Uncertainties Used for Probabilistic Reliability Assessments of Wave Energy Converters

DEFF Research Database (Denmark)

Ambühl, Simon; Kofoed, Jens Peter; Sørensen, John Dalsgaard

2015-01-01

Wave models used for site assessments are subjected to model uncertainties, which need to be quantified when using wave model results for probabilistic reliability assessments. This paper focuses on determination of wave model uncertainties. Four different wave models are considered, and validation...... data are collected from published scientific research. The bias and the root-mean-square error, as well as the scatter index, are considered for the significant wave height as well as the mean zero-crossing wave period. Based on an illustrative generic example, this paper presents how the quantified...... uncertainties can be implemented in probabilistic reliability assessments....

Determination of Wave Model Uncertainties used for Probabilistic Reliability Assessments of Wave Energy Devices

DEFF Research Database (Denmark)

Ambühl, Simon; Kofoed, Jens Peter; Sørensen, John Dalsgaard

2014-01-01

Wave models used for site assessments are subject to model uncertainties, which need to be quantified when using wave model results for probabilistic reliability assessments. This paper focuses on determination of wave model uncertainties. Considered are four different wave models and validation...... data is collected from published scientific research. The bias, the root-mean-square error as well as the scatter index are considered for the significant wave height as well as the mean zero-crossing wave period. Based on an illustrative generic example it is shown how the estimated uncertainties can...... be implemented in probabilistic reliability assessments....

On economic resolution and uncertainty in hydrocarbon exploration assessment

International Nuclear Information System (INIS)

Lerche, I.

1998-01-01

When assessment of parameters of a decision tree for a hydrocarbon exploration project can lie within estimated ranges, it is shown that the ensemble average expected value has two sorts of uncertainties: one is due to the expected value of each realization of the decision tree being different than the average; the second is due to intrinsic variance of each decision tree. The total standard error of the average expected value combines both sorts. The use of additional statistical measures, such as standard error, volatility, and cumulative probability of making a profit, provide insight into the selection process leading to a more appropriate decision. In addition, the use of relative contributions and relative importance for the uncertainty measures guides one to a better determination of those parameters that dominantly influence the total ensemble uncertainty. In this way one can concentrate resources on efforts to minimize the uncertainty ranges of such dominant parameters. A numerical illustration is provided to indicate how such calculations can be performed simply with a hand calculator. (author)

Summary of existing uncertainty methods

International Nuclear Information System (INIS)

Glaeser, Horst

2013-01-01

A summary of existing and most used uncertainty methods is presented, and the main features are compared. One of these methods is the order statistics method based on Wilks' formula. It is applied in safety research as well as in licensing. This method has been first proposed by GRS for use in deterministic safety analysis, and is now used by many organisations world-wide. Its advantage is that the number of potential uncertain input and output parameters is not limited to a small number. Such a limitation was necessary for the first demonstration of the Code Scaling Applicability Uncertainty Method (CSAU) by the United States Regulatory Commission (USNRC). They did not apply Wilks' formula in their statistical method propagating input uncertainties to obtain the uncertainty of a single output variable, like peak cladding temperature. A Phenomena Identification and Ranking Table (PIRT) was set up in order to limit the number of uncertain input parameters, and consequently, the number of calculations to be performed. Another purpose of such a PIRT process is to identify the most important physical phenomena which a computer code should be suitable to calculate. The validation of the code should be focused on the identified phenomena. Response surfaces are used in some applications replacing the computer code for performing a high number of calculations. The second well known uncertainty method is the Uncertainty Methodology Based on Accuracy Extrapolation (UMAE) and the follow-up method 'Code with the Capability of Internal Assessment of Uncertainty (CIAU)' developed by the University Pisa. Unlike the statistical approaches, the CIAU does compare experimental data with calculation results. It does not consider uncertain input parameters. Therefore, the CIAU is highly dependent on the experimental database. The accuracy gained from the comparison between experimental data and calculated results are extrapolated to obtain the uncertainty of the system code predictions

An introductory guide to uncertainty analysis in environmental and health risk assessment. Environmental Restoration Program

International Nuclear Information System (INIS)

Hammonds, J.S.; Hoffman, F.O.; Bartell, S.M.

1994-12-01

This report presents guidelines for evaluating uncertainty in mathematical equations and computer models applied to assess human health and environmental risk. Uncertainty analyses involve the propagation of uncertainty in model parameters and model structure to obtain confidence statements for the estimate of risk and identify the model components of dominant importance. Uncertainty analyses are required when there is no a priori knowledge about uncertainty in the risk estimate and when there is a chance that the failure to assess uncertainty may affect the selection of wrong options for risk reduction. Uncertainty analyses are effective when they are conducted in an iterative mode. When the uncertainty in the risk estimate is intolerable for decision-making, additional data are acquired for the dominant model components that contribute most to uncertainty. This process is repeated until the level of residual uncertainty can be tolerated. A analytical and numerical methods for error propagation are presented along with methods for identifying the most important contributors to uncertainty. Monte Carlo simulation with either Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) is proposed as the most robust method for propagating uncertainty through either simple or complex models. A distinction is made between simulating a stochastically varying assessment endpoint (i.e., the distribution of individual risks in an exposed population) and quantifying uncertainty due to lack of knowledge about a fixed but unknown quantity (e.g., a specific individual, the maximally exposed individual, or the mean, median, or 95%-tile of the distribution of exposed individuals). Emphasis is placed on the need for subjective judgement to quantify uncertainty when relevant data are absent or incomplete

Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

A Conceptual Methodology for Assessing Acquisition Requirements Robustness against Technology Uncertainties

Science.gov (United States)

Chou, Shuo-Ju

2011-12-01

In recent years the United States has shifted from a threat-based acquisition policy that developed systems for countering specific threats to a capabilities-based strategy that emphasizes the acquisition of systems that provide critical national defense capabilities. This shift in policy, in theory, allows for the creation of an "optimal force" that is robust against current and future threats regardless of the tactics and scenario involved. In broad terms, robustness can be defined as the insensitivity of an outcome to "noise" or non-controlled variables. Within this context, the outcome is the successful achievement of defense strategies and the noise variables are tactics and scenarios that will be associated with current and future enemies. Unfortunately, a lack of system capability, budget, and schedule robustness against technology performance and development uncertainties has led to major setbacks in recent acquisition programs. This lack of robustness stems from the fact that immature technologies have uncertainties in their expected performance, development cost, and schedule that cause to variations in system effectiveness and program development budget and schedule requirements. Unfortunately, the Technology Readiness Assessment process currently used by acquisition program managers and decision-makers to measure technology uncertainty during critical program decision junctions does not adequately capture the impact of technology performance and development uncertainty on program capability and development metrics. The Technology Readiness Level metric employed by the TRA to describe program technology elements uncertainties can only provide a qualitative and non-descript estimation of the technology uncertainties. In order to assess program robustness, specifically requirements robustness, against technology performance and development uncertainties, a new process is needed. This process should provide acquisition program managers and decision

Uncertainty characterization approaches for risk assessment of DBPs in drinking water: a review.

Science.gov (United States)

Chowdhury, Shakhawat; Champagne, Pascale; McLellan, P James

2009-04-01

The management of risk from disinfection by-products (DBPs) in drinking water has become a critical issue over the last three decades. The areas of concern for risk management studies include (i) human health risk from DBPs, (ii) disinfection performance, (iii) technical feasibility (maintenance, management and operation) of treatment and disinfection approaches, and (iv) cost. Human health risk assessment is typically considered to be the most important phase of the risk-based decision-making or risk management studies. The factors associated with health risk assessment and other attributes are generally prone to considerable uncertainty. Probabilistic and non-probabilistic approaches have both been employed to characterize uncertainties associated with risk assessment. The probabilistic approaches include sampling-based methods (typically Monte Carlo simulation and stratified sampling) and asymptotic (approximate) reliability analysis (first- and second-order reliability methods). Non-probabilistic approaches include interval analysis, fuzzy set theory and possibility theory. However, it is generally accepted that no single method is suitable for the entire spectrum of problems encountered in uncertainty analyses for risk assessment. Each method has its own set of advantages and limitations. In this paper, the feasibility and limitations of different uncertainty analysis approaches are outlined for risk management studies of drinking water supply systems. The findings assist in the selection of suitable approaches for uncertainty analysis in risk management studies associated with DBPs and human health risk.

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

International Nuclear Information System (INIS)

Chung, Dae Wook

2011-01-01

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

A reliability assessment methodology for the VHTR passive safety system

International Nuclear Information System (INIS)

Lee, Hyungsuk; Jae, Moosung

2014-01-01

The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

Comprehensive development plans for the low- and intermediate-level radioactive waste disposal facility in Korea and preliminary safety assessment

Energy Technology Data Exchange (ETDEWEB)

Jung, Kang Il; Kim, Jin Hyeong; Kwon, Mi Jin; Jeong, Mi Seon; Hong, Sung Wook; Park, Jin Beak [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

The disposal facility in Gyeongju is planning to dispose of 800,000 packages of low- and intermediate- level radioactive waste. This facility will be developed as a complex disposal facility that has various types of disposal facilities and accompanying management. In this study, based on the comprehensive development plan of the disposal facility, a preliminary post-closure safety assessment is performed to predict the phase development of the total capacity for the 800,000 packages to be disposed of at the site. The results for each scenario meet the performance target of the disposal facility. The assessment revealed that there is a significant impact of the inventory of intermediate-level radionuclide waste on the safety evaluation. Due to this finding, we introduce a disposal limit value for intermediate-level radioactive waste. With stepwise development of safety case, this development plan will increase the safety of disposal facilities by reducing uncertainties within the future development of the underground silo disposal facilities.

Preliminary Safety and Risk HSE Assessment. Application to the Potential Locations of a CO2 Geological Storage Pilot

International Nuclear Information System (INIS)

Recreo, F.; Eguilior, S.; Ruiz, C.; Lomba, L.; Hurtado, A.

2015-01-01

The location of a site safe and able to sequester CO2 for long periods of time is essential to gain public acceptance. This requires a long-term safety assessment developed in a robust and reliable framework. Site selection is the first step and requires specific research. This paper describes the application of the Selection and Classification Method of Geological Formations (SCF) developed to assess the potential of geological formations to CO2 storage. This assessment is based in the analysis of risks to Health, Safety and Environment (HSE) derived from potential CO2 leakage. Comparisons of the results obtained from a number of potential sites can help to select the best candidate for CO2 injection. The potential impact will be related to three key potential features of CO2 geological storage: the potential of the target geological formation for long term CO2 containment; the potential for secondary containment on containment failure of the target formation; and the site's potential to mitigate and/or disperse CO2 leakage if the primary and secondary containments fail. The methodology assesses each of these three characteristics through an analysis and assessment of properties of certain attributes of them. Uncertainty will remain as an input and output value of the methodology due to the usual lack of data in most site selection processes. The global uncertainty reports on the trust on the knowledge of the site characteristics. Therefore, the methodology enables comparing sites taking into account both the HSE risk expectation and the estimation of the quality of knowledge concerning such risk. The objective is to contribute to the selection of potential sites for a CO2 injection pilot plant in the Iberian Peninsula from the perspective of Safety and Risk Analysis.

Collaborative framework for PIV uncertainty quantification: comparative assessment of methods

International Nuclear Information System (INIS)

Sciacchitano, Andrea; Scarano, Fulvio; Neal, Douglas R; Smith, Barton L; Warner, Scott O; Vlachos, Pavlos P; Wieneke, Bernhard

2015-01-01

A posteriori uncertainty quantification of particle image velocimetry (PIV) data is essential to obtain accurate estimates of the uncertainty associated with a given experiment. This is particularly relevant when measurements are used to validate computational models or in design and decision processes. In spite of the importance of the subject, the first PIV uncertainty quantification (PIV-UQ) methods have been developed only in the last three years. The present work is a comparative assessment of four approaches recently proposed in the literature: the uncertainty surface method (Timmins et al 2012), the particle disparity approach (Sciacchitano et al 2013), the peak ratio criterion (Charonko and Vlachos 2013) and the correlation statistics method (Wieneke 2015). The analysis is based upon experiments conducted for this specific purpose, where several measurement techniques are employed simultaneously. The performances of the above approaches are surveyed across different measurement conditions and flow regimes. (paper)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Using sequential indicator simulation to assess the uncertainty of delineating heavy-metal contaminated soils

International Nuclear Information System (INIS)

Juang, Kai-Wei; Chen, Yue-Shin; Lee, Dar-Yuan

2004-01-01

Mapping the spatial distribution of soil pollutants is essential for delineating contaminated areas. Currently, geostatistical interpolation, kriging, is increasingly used to estimate pollutant concentrations in soils. The kriging-based approach, indicator kriging (IK), may be used to model the uncertainty of mapping. However, a smoothing effect is usually produced when using kriging in pollutant mapping. The detailed spatial patterns of pollutants could, therefore, be lost. The local uncertainty of mapping pollutants derived by the IK technique is referred to as the conditional cumulative distribution function (ccdf) for one specific location (i.e. single-location uncertainty). The local uncertainty information obtained by IK is not sufficient as the uncertainty of mapping at several locations simultaneously (i.e. multi-location uncertainty or spatial uncertainty) is required to assess the reliability of the delineation of contaminated areas. The simulation approach, sequential indicator simulation (SIS), which has the ability to model not only single, but also multi-location uncertainties, was used, in this study, to assess the uncertainty of the delineation of heavy metal contaminated soils. To illustrate this, a data set of Cu concentrations in soil from Taiwan was used. The results show that contour maps of Cu concentrations generated by the SIS realizations exhausted all the spatial patterns of Cu concentrations without the smoothing effect found when using the kriging method. Based on the SIS realizations, the local uncertainty of Cu concentrations at a specific location of x', refers to the probability of the Cu concentration z(x') being higher than the defined threshold level of contamination (z c ). This can be written as Prob SIS [z(x')>z c ], representing the probability of contamination. The probability map of Prob SIS [z(x')>z c ] can then be used for delineating contaminated areas. In addition, the multi-location uncertainty of an area A

Environmental impact and risk assessments and key factors contributing to the overall uncertainties.

Science.gov (United States)

Salbu, Brit

2016-01-01

There is a significant number of nuclear and radiological sources that have contributed, are still contributing, or have the potential to contribute to radioactive contamination of the environment in the future. To protect the environment from radioactive contamination, impact and risk assessments are performed prior to or during a release event, short or long term after deposition or prior and after implementation of countermeasures. When environmental impact and risks are assessed, however, a series of factors will contribute to the overall uncertainties. To provide environmental impact and risk assessments, information on processes, kinetics and a series of input variables is needed. Adding problems such as variability, questionable assumptions, gaps in knowledge, extrapolations and poor conceptual model structures, a series of factors are contributing to large and often unacceptable uncertainties in impact and risk assessments. Information on the source term and the release scenario is an essential starting point in impact and risk models; the source determines activity concentrations and atom ratios of radionuclides released, while the release scenario determine the physico-chemical forms of released radionuclides such as particle size distribution, structure and density. Releases will most often contain other contaminants such as metals, and due to interactions, contaminated sites should be assessed as a multiple stressor scenario. Following deposition, a series of stressors, interactions and processes will influence the ecosystem transfer of radionuclide species and thereby influence biological uptake (toxicokinetics) and responses (toxicodynamics) in exposed organisms. Due to the variety of biological species, extrapolation is frequently needed to fill gaps in knowledge e.g., from effects to no effects, from effects in one organism to others, from one stressor to mixtures. Most toxtests are, however, performed as short term exposure of adult organisms

Uncertainty assessment of urban pluvial flood risk in a context of climate change adaptation decision making

DEFF Research Database (Denmark)

Arnbjerg-Nielsen, Karsten; Zhou, Qianqian

2014-01-01

uncertainty analysis, which can assess and quantify the overall uncertainty in relation to climate change adaptation to urban flash floods. The analysis is based on an uncertainty cascade that by means of Monte Carlo simulations of flood risk assessments incorporates climate change impacts as a key driver......There has been a significant increase in climatic extremes in many regions. In Central and Northern Europe, this has led to more frequent and more severe floods. Along with improved flood modelling technologies this has enabled development of economic assessment of climate change adaptation...... to increasing urban flood risk. Assessment of adaptation strategies often requires a comprehensive risk-based economic analysis of current risk, drivers of change of risk over time, and measures to reduce the risk. However, such studies are often associated with large uncertainties. The uncertainties arise from...

From the feasibility assessment to the licensing application: organisation of the data acquisition, how to deal with metrological limits, uncertainties and project milestones; how far must we go?

International Nuclear Information System (INIS)

Landais, P.; Labalette, T.

2009-01-01

The research work summarised in the Dossier 2005 Argile has provided detailed information on each of the repository components but also on the determination, the analysis and the assessment of the main phenomena which are occurring within the repository. Their detailed representation associated with proposed repository architectures allowed the processing of the data in order to assess the robustness of the repository and to see how it would meet safety requirements. Through various indicators, the analysis showed that the three main safety functions preventing water circulation, limiting radionuclides release and immobilizing them in the repository and delaying and attenuating radionuclide migration were effectively fulfilled by the proposed system in both normal and much more penalizing situations. The PARS as well as the QSA already facilitated a systematic identification of uncertainties, then allowing covering them either trough cautious hypothesis, penalizing or conservative representation of some phenomena or components, sensitivity studies or altered evolution scenarios. Subsequently, the safety analysis revealed some residual uncertainties and margins for potential progress which will provide useful orientations for future research developments. While the safety analysis conducted reveals that the repository appears to be robust in all the configurations envisaged with respect to its safety functions, both CNE and safety authority evaluations focus on the necessity to provide more comprehensive and realistic modelling of the behavior of the repository (both exploitation and post-closure periods) and of the radionuclides. For example, it is requested not to consider the perturbed zone (EDZ) as a dead zone the characteristics and properties of which are set to zero in terms of transport. Similarly, when ANDRA constructed its safety case, an envelope hypothesis for conducting calculations led to consider the repository as fully saturated as soon as its

Procedures to relate the NII safety assessment principles for nuclear reactors to risk

CERN Document Server

Kelly, G N; Hemming, C R

1985-01-01

Within the framework of the Public Inquiry into the proposed pressurised water reactor (PWR) at Sizewell, estimates were made of the levels of individual and societal risk from a PWR designed in a manner which would conform to the safety assessment principles formulated by the Nuclear Installations Inspectorate (NII). The procedures used to derive these levels of risk are described in this report. The opportunity has also been taken to revise the risk estimates made at the time of the Inquiry by taking account of additional data which were not then available, and to provide further quantification of the likely range of uncertainty in the predictions. This re-analysis has led to small changes in the levels of risk previously evaluated, but these are not sufficient to affect the broad conclusions reached before. For a reactor just conforming to the NII safety assessment principles a maximum individual risk of fatal cancer of about 10 sup - sup 6 per year of reactor operation has been estimated; the societal ris...