WorldWideScience

Sample records for chemicals safety analysis

  1. Safety- and Risk Analysis Activities in Chemical Industry in Europe

    International Nuclear Information System (INIS)

    The current paper gives an overview of the legislation and the methods used in safety and risk management in the chemical industry within Europe and in particular within the European Union. The paper is based on a report that has been written for the SOS-1 project under the Nordic nuclear safety research (NKS). Safety- and risk-related matters in the process industry, in particular, in chemical, within the EU are subject to consideration at three levels: (1) EU legislation, (2) European/intemational standardisation, and (3) socio-economic analysis. EC Directives define the 'essential requirements', e.g., protection of health and safety, that must be fulfilled when goods are placed on the market or some industry is put into operation. The European standards bodies (CEN, CENELEC and ETSI) have the task of establishing the corresponding technical specifications, meeting the essential requirements of the Directives, compliance with which will provide a presumption of conformity with the essential requirements. Such specifications are referred to as 'harmonised standards'. Compliance with harmonised standards remains voluntary, and manufacturers are free to choose any other technical solution that provides compliance with the essential requirements. This view is stated in the 'New Approach' to technical harmonisation and standardisation (details can be found on the web page: http://europe.eu.int/comm/enterprise/newapproach/standardization/index .html). Standardisation as well as the regulation of technical risks is increasingly being undertaken at European or international level. The European legislator limits its role to the affirmation of overall objectives, and leaves it to the economic players to draw up the technical procedures and standards to specify in detail the ways and means of attaining them. Many countries have introduced requirements that new legislation and/or administrative regulations be subject to socio-economic analysis. In this respect there is a

  2. Safety- and Risk Analysis Activities in Chemical Industry in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Kozine, Igor; Duijm, Nijs Jan; Lauridsen Kurt [Risoe National Laboratory, Roskilde (Denmark). Systems Analysis Department

    2001-07-01

    The current paper gives an overview of the legislation and the methods used in safety and risk management in the chemical industry within Europe and in particular within the European Union. The paper is based on a report that has been written for the SOS-1 project under the Nordic nuclear safety research (NKS). Safety- and risk-related matters in the process industry, in particular, in chemical, within the EU are subject to consideration at three levels: (1) EU legislation, (2) European/intemational standardisation, and (3) socio-economic analysis. EC Directives define the 'essential requirements', e.g., protection of health and safety, that must be fulfilled when goods are placed on the market or some industry is put into operation. The European standards bodies (CEN, CENELEC and ETSI) have the task of establishing the corresponding technical specifications, meeting the essential requirements of the Directives, compliance with which will provide a presumption of conformity with the essential requirements. Such specifications are referred to as 'harmonised standards'. Compliance with harmonised standards remains voluntary, and manufacturers are free to choose any other technical solution that provides compliance with the essential requirements. This view is stated in the 'New Approach' to technical harmonisation and standardisation (details can be found on the web page: http://europe.eu.int/comm/enterprise/newapproach/standardization/index .html). Standardisation as well as the regulation of technical risks is increasingly being undertaken at European or international level. The European legislator limits its role to the affirmation of overall objectives, and leaves it to the economic players to draw up the technical procedures and standards to specify in detail the ways and means of attaining them. Many countries have introduced requirements that new legislation and/or administrative regulations be subject to socio-economic analysis

  3. Final Safety Analysis Document for Building 693 Chemical Waste Storage Building at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    This Safety Analysis Document (SAD) for the Lawrence Livermore National Laboratory (LLNL) Building 693, Chemical Waste Storage Building (desipated as Building 693 Container Storage Unit in the Laboratory's RCRA Part B permit application), provides the necessary information and analyses to conclude that Building 693 can be operated at low risk without unduly endangering the safety of the building operating personnel or adversely affecting the public or the environment. This Building 693 SAD consists of eight sections and supporting appendices. Section 1 presents a summary of the facility designs and operations and Section 2 summarizes the safety analysis method and results. Section 3 describes the site, the facility desip, operations and management structure. Sections 4 and 5 present the safety analysis and operational safety requirements (OSRs). Section 6 reviews Hazardous Waste Management's (HWM) Quality Assurance (QA) program. Section 7 lists the references and background material used in the preparation of this report Section 8 lists acronyms, abbreviations and symbols. Appendices contain supporting analyses, definitions, and descriptions that are referenced in the body of this report

  4. Do provisions to advance chemical facility safety also advance chemical facility security? An analysis of possible synergies

    DEFF Research Database (Denmark)

    Hedlund, Frank Huess

    2012-01-01

    The European Commission has launched a study on the applicability of existing chemical industry safety provisions to enhancing security of chemical facilities covering the situation in 18 EU Member States. This paper reports some preliminary analytical findings regarding the extent to which...... Infrastructures (ECI Directive) addresses facility security but does not cover the chemical sector. Chemical facility safety at EU level is addressed by way of the Seveso-II Directive. Preliminary estimates by the chemical industry suggest that perhaps 80% of the existing safety measures under Seveso-II would...... existing provisions that have been put into existence to advance safety objectives due to synergy effects could be expected advance security objectives as well. The paper provides a conceptual definition of safety and security and presents a framework of their essential components. Key differences are...

  5. Chemical Safety – Introduction

    CERN Multimedia

    DG Unit

    2009-01-01

    A course of "Chemical Safety – Introduction" will be held in English on 29 May 2009, 9:30-12:00. There are some places left. If you are interested in participating, please register on the Training Catalogue. You will then receive an invitation by email.

  6. Conventional and dynamic safety analysis: Comparison on a chemical batch reactor

    International Nuclear Information System (INIS)

    Dynamic safety analysis methodologies are an attractive approach to tackle systems with complex dynamics (i.e. with behavior highly dependent on the values of the process parameters): this is often the case in various areas of the chemical industry. The present paper compares analyses with Probabilistic Safety Assessment (PSA)/Quantitative Risk Assessment (QRA) methods with those from a dynamic methodology (Monte Carlo simulation). The results of a case study for a chemical batch reactor from the literature, overall risk figure and main contributors, are examined. The comparison has shown that, provided that the event success criteria are appropriately defined, consistent results can be obtained; otherwise important accident scenarios, identifiable by the dynamic Monte Carlo simulation, are possibly missed in the application of conventional methods. Defining such criteria was quite resource-intensive: for the analysis of this small system, the success criteria definitions required many system simulation runs (about 1000). Such large numbers of runs may not be practical in industrial-scale applications. It is shown that success criteria obtained with fewer simulation runs could have led to different quantitative PSA results and to the omission of important accident scenario variants.

  7. Guidance on health effects of toxic chemicals. Safety Analysis Report Update Program

    Energy Technology Data Exchange (ETDEWEB)

    Foust, C.B.; Griffin, G.D.; Munro, N.B.; Socolof, M.L.

    1994-02-01

    Martin Marietta Energy Systems, Inc. (MMES), and Martin Marietta Utility Services, Inc. (MMUS), are engaged in phased programs to update the safety documentation for the existing US Department of Energy (DOE)-owned facilities. The safety analysis of potential toxic hazards requires a methodology for evaluating human health effects of predicted toxic exposures. This report provides a consistent set of health effects and documents toxicity estimates corresponding to these health effects for some of the more important chemicals found within MMES and MMUS. The estimates are based on published toxicity information and apply to acute exposures for an ``average`` individual. The health effects (toxicological endpoints) used in this report are (1) the detection threshold; (2) the no-observed adverse effect level; (3) the onset of irritation/reversible effects; (4) the onset of irreversible effects; and (5) a lethal exposure, defined to be the 50% lethal level. An irreversible effect is defined as a significant effect on a person`s quality of life, e.g., serious injury. Predicted consequences are evaluated on the basis of concentration and exposure time.

  8. Do provisions to advance chemical facility safety also advance chemical facility security? An analysis of possible synergies

    DEFF Research Database (Denmark)

    Hedlund, Frank Huess

    The European Commission has launched a study on the applicability of existing chemical industry safety provisions to enhancing security of chemical facilities covering the situation in 18 EU Member States. This paper reports some preliminary analytical findings regarding the extent to which...... exist at the mitigation level. At the strategic policy level, synergies are obvious. The security of chemical facilities is important. First, facilities with large inventories of toxic materials could be attractive targets for terrorists. The concern is sabotage causing an intentional release that could...

  9. Chemical process hazards analysis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    The Office of Worker Health and Safety (EH-5) under the Assistant Secretary for the Environment, Safety and Health of the US Department (DOE) has published two handbooks for use by DOE contractors managing facilities and processes covered by the Occupational Safety and Health Administration (OSHA) Rule for Process Safety Management of Highly Hazardous Chemicals (29 CFR 1910.119), herein referred to as the PSM Rule. The PSM Rule contains an integrated set of chemical process safety management elements designed to prevent chemical releases that can lead to catastrophic fires, explosions, or toxic exposures. The purpose of the two handbooks, ``Process Safety Management for Highly Hazardous Chemicals`` and ``Chemical Process Hazards Analysis,`` is to facilitate implementation of the provisions of the PSM Rule within the DOE. The purpose of this handbook ``Chemical Process Hazards Analysis,`` is to facilitate, within the DOE, the performance of chemical process hazards analyses (PrHAs) as required under the PSM Rule. It provides basic information for the performance of PrHAs, and should not be considered a complete resource on PrHA methods. Likewise, to determine if a facility is covered by the PSM rule, the reader should refer to the handbook, ``Process Safety Management for Highly Hazardous Chemicals`` (DOE- HDBK-1101-96). Promulgation of the PSM Rule has heightened the awareness of chemical safety management issues within the DOE. This handbook is intended for use by DOE facilities and processes covered by the PSM rule to facilitate contractor implementation of the PrHA element of the PSM Rule. However, contractors whose facilities and processes not covered by the PSM Rule may also use this handbook as a basis for conducting process hazards analyses as part of their good management practices. This handbook explains the minimum requirements for PrHAs outlined in the PSM Rule. Nowhere have requirements been added beyond what is specifically required by the rule.

  10. Chemical Hygiene and Safety Plan

    Energy Technology Data Exchange (ETDEWEB)

    Berkner, K.

    1992-08-01

    The objective of this Chemical Hygiene and Safety Plan (CHSP) is to provide specific guidance to all LBL employees and contractors who use hazardous chemicals. This Plan, when implemented, fulfills the requirements of both the Federal OSHA Laboratory Standard (29 CFR 1910.1450) for laboratory workers, and the Federal OSHA Hazard Communication Standard (29 CFR 1910.1200) for non-laboratory operations (e.g., shops). It sets forth safety procedures and describes how LBL employees are informed about the potential chemical hazards in their work areas so they can avoid harmful exposures and safeguard their health. Generally, communication of this Plan will occur through training and the Plan will serve as a the framework and reference guide for that training.

  11. An analysis of the chemical safety of secondary effluent for reuse purposes and the requirement for advanced treatment.

    Science.gov (United States)

    Jin, Pengkang; Jin, Xin; Wang, Xiaochang C; Shi, Xinbin

    2013-04-01

    This paper presents a study on the chemical safety of the secondary effluent for reuse purposes and the requirement of advanced treatment. Water quality analysis was conducted regarding conventional chemical items, hazardous metals, trace organics and endocrine disrupting chemicals (EDCs). Generally speaking, the turbidity, COD, BOD, TN and TP of the secondary effluent can meet the Chinese standards for urban miscellaneous water reuse but higher colour is a problem. Further removal of BOD and TP may still be required if the water is reused for landscape and environmental purposes especially relating to recreation. In addition, Hazardous metals, trace organics and endocrine disrupting chemicals (EDCs) are not the main problems for water reuse. At the same time, several tertiary treatment processes were evaluated. The coagulation-filtration process is effective process for further improvement of the conventional water quality items and removal of hazardous metals but less effective in dealing with dissolved organic matter. The ultrafiltration (UF) can achieve almost complete removal of turbid matter while its ability to remove dissolved substances is limited. The ozone-biofiltration is the most effective for colour and organic removal but it can hardly remove the residual hazardous metals. Therefore, the selection of suitable process for different water quality is important for water use. PMID:23384543

  12. Auditable safety analysis: High Radiation Level Chemical Development Facility (Buildings 4507 and 4556), Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    The High-Radiation-Level Chemical Development Facility includes Buildings 4507 and 4556. Building 4507, located immediately to the west of Building 4500N and to the south of Building 4505, is a doubly contained three-level structure constructed in 1957. The most recent use of the facility was for recovery of multi-gram quantities of 244Cm during the early 1970s and for Liquid Metal Fast Breeder Reactor (LMFBR) fuel studies in the late 1970s. It has remained in safe standby since 1980. Building 4556 is a below-grade filter pit located to the southwest of Building 4507 and was constructed in 1972. Ventilation from the cells in Building 4507 is passed through high-efficiency particulate air (HEPA) filtration in this building prior to being exhausted to the Building 3039 stack system. This building remains in operation to support ventilation requirements for Building 4507. This Auditable Safety Analysis (ASA) was developed in accordance with the requirements in Energy Systems Program Description FS-103PD, Safety Documentation, Revision 1. This ASA identifies and screens all hazards associated with Buildings 4507 and 4556. The only hazard not screened out and requiring further analysis following the initial screening process is radioactive material in the form of surface contamination. The results of this ASA indicate that the hazards associated with Buildings 4507 and 4556 do not pose a significant threat to workers, the public, or the environment

  13. New set of Chemical Safety rules

    CERN Multimedia

    HSE Unit

    2011-01-01

    A new set of four Safety Rules was issued on 28 March 2011: Safety Regulation SR-C ver. 2, Chemical Agents (en); General Safety Instruction GSI-C1, Prevention and Protection Measures (en); General Safety Instruction GSI-C2, Explosive Atmospheres (en); General Safety Instruction GSI-C3, Monitoring of Exposure to Hazardous Chemical Agents in Workplace Atmospheres (en). These documents form part of the CERN Safety Rules and are issued in application of the “Staff Rules and Regulations” and of document SAPOCO 42. These documents set out the minimum requirements for the protection of persons from risks to their occupational safety and health arising, or likely to arise, from the effects of hazardous chemical agents that are present in the workplace or used in any CERN activity. Simultaneously, the HSE Unit has published seven Safety Guidelines and six Safety Forms. These documents are available from the dedicated Web page “Chemical, Cryogenic and Biological Safety&...

  14. Chemical Safety Vulnerability Working Group Report

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    This report marks the culmination of a 4-month review conducted to identify chemical safety vulnerabilities existing at DOE facilities. This review is an integral part of DOE's efforts to raise its commitment to chemical safety to the same level as that for nuclear safety.

  15. Experiments To Demonstrate Chemical Process Safety Principles.

    Science.gov (United States)

    Dorathy, Brian D.; Mooers, Jamisue A.; Warren, Matthew M.; Mich, Jennifer L.; Murhammer, David W.

    2001-01-01

    Points out the need to educate undergraduate chemical engineering students on chemical process safety and introduces the content of a chemical process safety course offered at the University of Iowa. Presents laboratory experiments demonstrating flammability limits, flash points, electrostatic, runaway reactions, explosions, and relief design.…

  16. New Safety rule for Chemical Agents

    CERN Multimedia

    Safety Commission

    2010-01-01

    The following Safety rule has been issued on 08-01-2010: Safety Regulation SR-C Chemical Agents This document applies to all persons under the Director General’s authority. It sets out the minimal requirements for the protection of persons from risks to their safety and health arising, or likely to arise, from the effects of hazardous chemical agents used in any CERN activity. All Safety rules are available on the web pages.

  17. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

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

  18. Microprocessors in automatic chemical analysis

    International Nuclear Information System (INIS)

    Application of microprocessors to programming and computing of solutions chemical analysis by a sequential technique is examined. Safety, performances reliability are compared to other methods. An example is given on uranium titration by spectrophotometry

  19. Chemical analysis of estragole in fennel based teas and associated safety assessment using the Margin of Exposure (MOE) approach

    NARCIS (Netherlands)

    Berg, van den S.J.P.L.; Alhusainy, W.; Restani, P.; Rietjens, I.

    2014-01-01

    This study describes the analysis of estragole in dry fennel preparations and in infusions prepared from them and an associated safety assessment. A wide range of estragole levels of 0.15–13.3 mg/g dry fennel preparation was found. The estragole content in infusions was considerably lower ranging be

  20. Reactor Safety Analysis

    International Nuclear Information System (INIS)

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

  1. K Basin safety analysis

    International Nuclear Information System (INIS)

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

  2. K Basin safety analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-16

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

  3. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

    Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...... the same system model and that this model is formalized in a real-time, interval logic, based on a conventional dynamic systems model with a state over time. The three safety analysis techniques are interpreted in this model and it is shown how to derive safety requirements for components of a system....

  4. Safety Considerations in the Chemical Process Industries

    Science.gov (United States)

    Englund, Stanley M.

    There is an increased emphasis on chemical process safety as a result of highly publicized accidents. Public awareness of these accidents has provided a driving force for industry to improve its safety record. There has been an increasing amount of government regulation.

  5. Safety analysis of NPP

    International Nuclear Information System (INIS)

    This paper presents a short review of the parallel safety analysis of the various types of NPP. The NPP with PWR, WWER, BWR and HWR type reactors are mentioned. Technical, economic, location and ecology aspects of the safety of the NPP have been analysed. (author)

  6. Safety analysis for `Fugen`

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

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

  7. Reactor Safety Analysis

    International Nuclear Information System (INIS)

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

  8. Chemical Security Analysis Center

    Data.gov (United States)

    Federal Laboratory Consortium — In 2006, by Presidential Directive, DHS established the Chemical Security Analysis Center (CSAC) to identify and assess chemical threats and vulnerabilities in the...

  9. Database for Safety-Oriented Tracking of Chemicals

    Science.gov (United States)

    Stump, Jacob; Carr, Sandra; Plumlee, Debrah; Slater, Andy; Samson, Thomas M.; Holowaty, Toby L.; Skeete, Darren; Haenz, Mary Alice; Hershman, Scot; Raviprakash, Pushpa

    2010-01-01

    SafetyChem is a computer program that maintains a relational database for tracking chemicals and associated hazards at Johnson Space Center (JSC) by use of a Web-based graphical user interface. The SafetyChem database is accessible to authorized users via a JSC intranet. All new chemicals pass through a safety office, where information on hazards, required personal protective equipment (PPE), fire-protection warnings, and target organ effects (TOEs) is extracted from material safety data sheets (MSDSs) and recorded in the database. The database facilitates real-time management of inventory with attention to such issues as stability, shelf life, reduction of waste through transfer of unused chemicals to laboratories that need them, quantification of chemical wastes, and identification of chemicals for which disposal is required. Upon searching the database for a chemical, the user receives information on physical properties of the chemical, hazard warnings, required PPE, a link to the MSDS, and references to the applicable International Standards Organization (ISO) 9000 standard work instructions and the applicable job hazard analysis. Also, to reduce the labor hours needed to comply with reporting requirements of the Occupational Safety and Health Administration, the data can be directly exported into the JSC hazardous- materials database.

  10. Process safety management for highly hazardous chemicals

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Purpose of this document is to assist US DOE contractors who work with threshold quantities of highly hazardous chemicals (HHCs), flammable liquids or gases, or explosives in successfully implementing the requirements of OSHA Rule for Process Safety Management of Highly Hazardous Chemicals (29 CFR 1910.119). Purpose of this rule is to prevent releases of HHCs that have the potential to cause catastrophic fires, explosions, or toxic exposures.

  11. Idaho Chemical Processing Plant safety document ICPP hazardous chemical evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Harwood, B.J.

    1993-01-01

    This report presents the results of a hazardous chemical evaluation performed for the Idaho Chemical Processing Plant (ICPP). ICPP tracks chemicals on a computerized database, Haz Track, that contains roughly 2000 individual chemicals. The database contains information about each chemical, such as its form (solid, liquid, or gas); quantity, either in weight or volume; and its location. The Haz Track database was used as the primary starting point for the chemical evaluation presented in this report. The chemical data and results presented here are not intended to provide limits, but to provide a starting point for nonradiological hazards analysis.

  12. Computer aided safety analysis

    International Nuclear Information System (INIS)

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

  13. Software safety hazard analysis

    International Nuclear Information System (INIS)

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

  14. Animal-Free Chemical Safety Assessment.

    Science.gov (United States)

    Loizou, George D

    2016-01-01

    The exponential growth of the Internet of Things and the global popularity and remarkable decline in cost of the mobile phone is driving the digital transformation of medical practice. The rapidly maturing digital, non-medical world of mobile (wireless) devices, cloud computing and social networking is coalescing with the emerging digital medical world of omics data, biosensors and advanced imaging which offers the increasingly realistic prospect of personalized medicine. Described as a potential "seismic" shift from the current "healthcare" model to a "wellness" paradigm that is predictive, preventative, personalized and participatory, this change is based on the development of increasingly sophisticated biosensors which can track and measure key biochemical variables in people. Additional key drivers in this shift are metabolomic and proteomic signatures, which are increasingly being reported as pre-symptomatic, diagnostic and prognostic of toxicity and disease. These advancements also have profound implications for toxicological evaluation and safety assessment of pharmaceuticals and environmental chemicals. An approach based primarily on human in vivo and high-throughput in vitro human cell-line data is a distinct possibility. This would transform current chemical safety assessment practice which operates in a human "data poor" to a human "data rich" environment. This could also lead to a seismic shift from the current animal-based to an animal-free chemical safety assessment paradigm. PMID:27493630

  15. Animal-Free Chemical Safety Assessment

    Science.gov (United States)

    Loizou, George D.

    2016-01-01

    The exponential growth of the Internet of Things and the global popularity and remarkable decline in cost of the mobile phone is driving the digital transformation of medical practice. The rapidly maturing digital, non-medical world of mobile (wireless) devices, cloud computing and social networking is coalescing with the emerging digital medical world of omics data, biosensors and advanced imaging which offers the increasingly realistic prospect of personalized medicine. Described as a potential “seismic” shift from the current “healthcare” model to a “wellness” paradigm that is predictive, preventative, personalized and participatory, this change is based on the development of increasingly sophisticated biosensors which can track and measure key biochemical variables in people. Additional key drivers in this shift are metabolomic and proteomic signatures, which are increasingly being reported as pre-symptomatic, diagnostic and prognostic of toxicity and disease. These advancements also have profound implications for toxicological evaluation and safety assessment of pharmaceuticals and environmental chemicals. An approach based primarily on human in vivo and high-throughput in vitro human cell-line data is a distinct possibility. This would transform current chemical safety assessment practice which operates in a human “data poor” to a human “data rich” environment. This could also lead to a seismic shift from the current animal-based to an animal-free chemical safety assessment paradigm. PMID:27493630

  16. Safety assessment principles for nuclear chemical plant

    International Nuclear Information System (INIS)

    The subject is covered in sections, as follows: foreword; introduction (functions of Health and Safety Executive and Nuclear Installations Inspectorate; scope of document and principles); fundamental requirements and policy; basic principles (radiological principles; principles for the evaluation of (a) radiation exposures under normal operating conditions, and (b) fault conditions and protection systems); engineering principles (general; radioactive materials control; movement of radioactive materials; radioactive waste and scrap control; radiological protection practice; protection systems; essential resources; plant containment and ventilation; plant operation; analysis of plant faults, transients and abnormal conditions; reliability analysis; external hazards; layout; installation checks and commissioning; servicing; decommissioning); management principles (the management of safety; quality assurance). (U.K.)

  17. Safety analysis and review system

    International Nuclear Information System (INIS)

    Westinghouse Savannah River Company (WSRC) has developed a comprehensive Safety Analysis and Review System that satisfies Department of Energy safety analysis report requirements. This system consists of interrelated criteria for hazard classification, risk assessment, selection of Safety Class Items (SCIs), and selection of Operational Safety Requirements (OSRs). The system provides input for design decisions at appropriate project milestones as required by the life cycle of a project. The criteria used for selection in hazard classification, risk assessment, Safety Class Items (SCI) identification, and Operational Safety Requirement (OSR) identification are the subject of this paper

  18. Preclosure Safety Analysis Guide

    International Nuclear Information System (INIS)

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

  19. Chemical exchange program analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Waffelaert, Pascale

    2007-09-01

    As part of its EMS, Sandia performs an annual environmental aspects/impacts analysis. The purpose of this analysis is to identify the environmental aspects associated with Sandia's activities, products, and services and the potential environmental impacts associated with those aspects. Division and environmental programs established objectives and targets based on the environmental aspects associated with their operations. In 2007 the most significant aspect identified was Hazardous Materials (Use and Storage). The objective for Hazardous Materials (Use and Storage) was to improve chemical handling, storage, and on-site movement of hazardous materials. One of the targets supporting this objective was to develop an effective chemical exchange program, making a business case for it in FY07, and fully implementing a comprehensive chemical exchange program in FY08. A Chemical Exchange Program (CEP) team was formed to implement this target. The team consists of representatives from the Chemical Information System (CIS), Pollution Prevention (P2), the HWMF, Procurement and the Environmental Management System (EMS). The CEP Team performed benchmarking and conducted a life-cycle analysis of the current management of chemicals at SNL/NM and compared it to Chemical Exchange alternatives. Those alternatives are as follows: (1) Revive the 'Virtual' Chemical Exchange Program; (2) Re-implement a 'Physical' Chemical Exchange Program using a Chemical Information System; and (3) Transition to a Chemical Management Services System. The analysis and benchmarking study shows that the present management of chemicals at SNL/NM is significantly disjointed and a life-cycle or 'Cradle-to-Grave' approach to chemical management is needed. This approach must consider the purchasing and maintenance costs as well as the cost of ultimate disposal of the chemicals and materials. A chemical exchange is needed as a mechanism to re-apply chemicals on site. This

  20. Probabilistic safety assessment in the chemical and nuclear industries

    CERN Document Server

    Fullwood, Ralph R

    2000-01-01

    Probabilistic Safety Analysis (PSA) determines the probability and consequences of accidents, hence, the risk. This subject concerns policy makers, regulators, designers, educators and engineers working to achieve maximum safety with operational efficiency. Risk is analyzed using methods for achieving reliability in the space program. The first major application was to the nuclear power industry, followed by applications to the chemical industry. It has also been applied to space, aviation, defense, ground, and water transportation. This book is unique in its treatment of chemical and nuclear risk. Problems are included at the end of many chapters, and answers are in the back of the book. Computer files are provided (via the internet), containing reliability data, a calculator that determines failure rate and uncertainty based on field experience, pipe break calculator, event tree calculator, FTAP and associated programs for fault tree analysis, and a units conversion code. It contains 540 references and many...

  1. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

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

  2. Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

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

  3. ILO activities in the area of chemical safety.

    Science.gov (United States)

    Obadia, Isaac

    2003-08-21

    The ILO has been active in the area of safety in the use of chemicals at work since the year of its creation in 1919, including the development of international treaties and other technical instruments, the provision of technical assistance to its member States, and the development of chemical safety information systems. The two key ILO standards in this area are the Conventions on safety in the use of chemicals at work (No. 170, 1990), and the Prevention of Major Industrial Accidents (No. 174, 1993). The ILO Programme on occupational safety, health and environment (Safe Work) is currently responsible for ILO chemical safety activities. In the past two decades, most of ILO work in this area has been carried out within the context of inter-agency collaboration frameworks linking the ILO, WHO, UNEP, FAO, UNIDO, UNITAR, and the OECD, including the International Programme on Chemical Safety (IPCS), the Inter-Organisation Programme for the Sound Management of Chemicals (IOMC), and the Intergovernmental Forum on Chemical Safety (IFCS). Apart from the regular development, updating and dissemination of chemical safety information data bases such as the IPCS International Chemical Cards, the elaboration of a Globally harmonized system for the classification and labelling of Chemicals (GHS) has been the most outstanding achievement of this international collaboration on chemical safety. PMID:12909402

  4. Meeting on risk and monitoring analysis techniques for food safety - RLA/5/060/ARCAL Project (ARCAL CXXVIII): sampling plans and introduction to chemical risk assessment in food innocuousness

    International Nuclear Information System (INIS)

    Some of the Latinoamerican countries such us Bolivia, Colombia, Uruguay and Venezuela participant in the meeting gave an exposition about the risk analysis and monitoring techniques in food safety in their countyries. With the aim to study components of risk analysis, food innocuousness, evaluation and chemical dangers, toxicity, exposure, change of paradigms in the global food system, data sources, study in animals and in vitro, sensitivity analysis, risk assessment in health it carried out the meeting

  5. Safety criteria for nuclear chemical plants

    International Nuclear Information System (INIS)

    Safety measures have always been required to limit the hazards due to accidental release of radioactive substances from nuclear power plants and chemical plants. The risk associated with the discharge of radioactive substances during normal operation has also to be kept acceptably low. BNFL (British Nuclear Fuels Ltd.) are developing risk criteria as targets for safe plant design and operation. The numerical values derived are compared with these criteria to see if plants are 'acceptably safe'. However, the criteria are not mandatory and may be exceeded if this can be justified. The risk assessments are subject to independent review and audit. The Nuclear Installations Inspectorate also has to pass the plants as safe. The assessment principles it uses are stated. The development of risk criteria for a multiplant site (nuclear chemical plants tend to be sited with many others which are related functionally) is discussed. This covers individual members of the general public, societal risks, risks to the workforce and external hazards. (U.K.)

  6. The chemical safety of irradiated foods

    International Nuclear Information System (INIS)

    While animal feeding studies and other biological testing methods have contributed greatly to the establishment of the toxicological safety of irradiated foods, probably no other single factor has lent itself so conclusively to this end as the availability of an unprecedented volume of analytical chemistry data on radiolytic products generated in a variety of foods and their raw materials and ingredients, collected at laboratories worldwide over decades. Such direct analytical chemical evidence, backed up by a general knowledge of radiation chemistry of bio-organic materials has allowed regulatory scientists and other competent, qualified and objective interested parties to discern with a high degree of confidence what takes place chemically at the sub-molecular level, and in the parts-per-trillion range, as a result of food irradiation. Ironically, this has also opened the way for nonqualified, subjectively negatively biased individuals to, for example, grossly misrepresent such compounds as benzene and formaldehyde in this context in an alarmist fashion to anyone predisposed to listen

  7. Safety analysis procedures for PHWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-01

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

  8. Views on chemical safety information and influences on chemical disposal behaviour in the UK

    International Nuclear Information System (INIS)

    This study examined how groups representing four tiers in the chemical supply chain (manufacturers, vendors, workers and consumers) understood safety information, and the factors that influenced disposal behaviour. Data from seven, semi-structured, focus groups was analysed both qualitatively (textual analysis) and quantitatively (network analysis). Such combined analytical methods enabled us to achieve both detailed insights into perceptions and behaviour and an objective understanding of the prevailing opinions that occurred within and between the focus group discussions. We found issues around awareness, trust, access and disposal behaviours differed between groups within the supply chain. Participants from the lower tiers perceived chemical safety information to be largely inaccessible. Labels were the main source of information on chemical risks for the middle and bottom tiers of the supply chain. Almost all of the participants were aware of the St Andrew's Cross and skull and crossbones symbols but few were familiar with the Volatile Organic Compound logo or the fish and tree symbol. Both the network and thematic analysis demonstrated that whilst frequent references to health risks associated with chemicals were made environmental risks were usually only articulated after prompting. It is clear that the issues surrounding public understanding of chemical safety labels are highly complex and this is compounded by inconsistencies in the cognitive profiles of chemical users. Substantially different cognitive profiles are likely to contribute towards communication difficulties between different tiers of the supply chain. Further research is needed to examine the most effective ways of communicating chemical hazards information to the public. The findings demonstrate a need to improve and simplify disposal guidance to members of the public, to raise public awareness of the graphic symbols in the CHIP 3.1, 2005 regulations and to improve access to disposal guidance

  9. Views on chemical safety information and influences on chemical disposal behaviour in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Hinks, J. [Enviresearch Ltd., Nanotechnology Centre, Herschel Building, Newcastle University, Newcastle upon Tyne, NE1 7RU (United Kingdom); Bush, J. [Institute for Health and Society, Newcastle University, William Leech Building, NE2 4HH (United Kingdom)], E-mail: Judith.bush@ncl.ac.uk; Andras, P. [School of Computing Science, Newcastle University, Claremont Tower, Newcastle University, NE1 7RU (United Kingdom); Garratt, J. [Institute for Health and Society, Newcastle University, William Leech Building, NE2 4HH (United Kingdom); Pigott, G. [NuFarm UK, Wyke, Bradford, West Yorkshire, BD12 9EJ (United Kingdom); Kennedy, A. [Enviresearch Ltd., Nanotechnology Centre, Herschel Building, Newcastle University, Newcastle upon Tyne, NE1 7RU (United Kingdom); Pless-Mulloli, T. [Institute for Health and Society, Newcastle University, William Leech Building, NE2 4HH (United Kingdom)

    2009-02-01

    This study examined how groups representing four tiers in the chemical supply chain (manufacturers, vendors, workers and consumers) understood safety information, and the factors that influenced disposal behaviour. Data from seven, semi-structured, focus groups was analysed both qualitatively (textual analysis) and quantitatively (network analysis). Such combined analytical methods enabled us to achieve both detailed insights into perceptions and behaviour and an objective understanding of the prevailing opinions that occurred within and between the focus group discussions. We found issues around awareness, trust, access and disposal behaviours differed between groups within the supply chain. Participants from the lower tiers perceived chemical safety information to be largely inaccessible. Labels were the main source of information on chemical risks for the middle and bottom tiers of the supply chain. Almost all of the participants were aware of the St Andrew's Cross and skull and crossbones symbols but few were familiar with the Volatile Organic Compound logo or the fish and tree symbol. Both the network and thematic analysis demonstrated that whilst frequent references to health risks associated with chemicals were made environmental risks were usually only articulated after prompting. It is clear that the issues surrounding public understanding of chemical safety labels are highly complex and this is compounded by inconsistencies in the cognitive profiles of chemical users. Substantially different cognitive profiles are likely to contribute towards communication difficulties between different tiers of the supply chain. Further research is needed to examine the most effective ways of communicating chemical hazards information to the public. The findings demonstrate a need to improve and simplify disposal guidance to members of the public, to raise public awareness of the graphic symbols in the CHIP 3.1, 2005 regulations and to improve access to disposal

  10. Using game theory to improve safety within chemical industrial parks

    CERN Document Server

    Reniers, Genserik

    2013-01-01

    Though the game-theoretic approach has been vastly studied and utilized in relation to economics of industrial organizations, it has hardly been used to tackle safety management in multi-plant chemical industrial settings. Using Game Theory for Improving Safety within Chemical Industrial Parks presents an in-depth discussion of game-theoretic modelling which may be applied to improve cross-company prevention and -safety management in a chemical industrial park.   By systematically analyzing game-theoretic models and approaches in relation to managing safety in chemical industrial parks, Using Game Theory for Improving Safety within Chemical Industrial Parks explores the ways game theory can predict the outcome of complex strategic investment decision making processes involving several adjacent chemical plants. A number of game-theoretic decision models are discussed to provide strategic tools for decision-making situations.   Offering clear and straightforward explanations of methodologies, Using Game Theor...

  11. Assessing food safety concepts on the dairy farm: the case of chemical hazards

    NARCIS (Netherlands)

    Valeeva, N.I.; Meuwissen, M.P.M.; Oude Lansink, A.G.J.M.; Bergevoet, R.H.M.; Huirne, R.B.M.

    2004-01-01

    Adaptive conjoint analysis was used to elicit farmers' and experts' preferences for attributes of improving food safety with respect to chemical hazards on the dairy farm. Groups of respondents were determined by cluster analysis based on similar farmers' and experts' perceptions of food safety impr

  12. Chemical Safety Vulnerability Working Group report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains the Executive summary; Introduction; Summary of vulnerabilities; Management systems weaknesses; Commendable practices; Summary of management response plan; Conclusions; and a Glossary of chemical terms.

  13. Chemical Safety Vulnerability Working Group report. Volume 1

    International Nuclear Information System (INIS)

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains the Executive summary; Introduction; Summary of vulnerabilities; Management systems weaknesses; Commendable practices; Summary of management response plan; Conclusions; and a Glossary of chemical terms

  14. Risk analysis and safety rationale

    International Nuclear Information System (INIS)

    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

  15. Chemical Safety Vulnerability Working Group report. Volume 3

    International Nuclear Information System (INIS)

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 3 consists of eleven appendices containing the following: Field verification reports for Idaho National Engineering Lab., Rocky Flats Plant, Brookhaven National Lab., Los Alamos National Lab., and Sandia National Laboratories (NM); Mini-visits to small DOE sites; Working Group meeting, June 7--8, 1994; Commendable practices; Related chemical safety initiatives at DOE; Regulatory framework and industry initiatives related to chemical safety; and Chemical inventory data from field self-evaluation reports

  16. Chemical Safety Vulnerability Working Group report. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 3 consists of eleven appendices containing the following: Field verification reports for Idaho National Engineering Lab., Rocky Flats Plant, Brookhaven National Lab., Los Alamos National Lab., and Sandia National Laboratories (NM); Mini-visits to small DOE sites; Working Group meeting, June 7--8, 1994; Commendable practices; Related chemical safety initiatives at DOE; Regulatory framework and industry initiatives related to chemical safety; and Chemical inventory data from field self-evaluation reports.

  17. Chemical Safety and Scientific Ethics in a Sophomore Chemistry Seminar

    Science.gov (United States)

    Moody, Anne E.; Griffith Freeman, R.

    1999-09-01

    A description of a course on chemical safety and scientific ethics is presented. The goals of this course are to impress upon the students the importance of safety in their professional lives; to empower the students to take charge of their own personal safety when working with chemicals; to illustrate and emphasize the vital importance of honesty and integrity within the scientific enterprise; and to explore issues of honesty and integrity through case studies that allow ethical decisions to be critically examined. The recent approaches and activities used to accomplish these goals are detailed. These include readings from chemical safety textbooks, chemical safety reports from news sources, and group discussions springing from problems in scientific ethics.

  18. Radwaste Disposal Safety Analysis

    International Nuclear Information System (INIS)

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

  19. Safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

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

  20. Radiometric chemical analysis

    International Nuclear Information System (INIS)

    The radiometric method of analysis is noted for its sensitivity and its simplicity in both apparatus and procedure. A few inexpensive radioactive reagents permit the analysis of a wide variety of chemical elements and compounds. Any particular procedure is generally applicable over a very wide range of concentrations. It is potentially an analytical method of great industrial significance. Specific examples of analyses are cited to illustrate the potentialities of ordinary equipment. Apparatus specifically designed for radiometric chemistry may shorten the time required, and increase the precision and accuracy for routine analyses. A sensitive and convenient apparatus for the routine performance of radiometric chemical analysis is a special type of centrifuge which has been used in obtaining the data presented in this paper. The radioactivity of the solution is measured while the centrifuge is spinning. This device has been used as the basis for an automatic analyser for phosphate ion, programmed to follow a sequence of unknown sampling, reagent mixing, centrifugation, counting data presentation, and phosphate replenishment. This analyser can repeatedly measure phosphate-concentration in the range of 5 to 50 ppm with an accuracy of ±5%. (author)

  1. Computer aided safety analysis 1989

    International Nuclear Information System (INIS)

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

  2. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    International Nuclear Information System (INIS)

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

  3. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    E.N. Lindner

    2004-12-03

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

  4. Safety analysis and risk assessment handbook

    International Nuclear Information System (INIS)

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

  5. Adapting safety requirements analysis to intrusion detection

    Science.gov (United States)

    Lutz, R.

    2001-01-01

    Several requirements analysis techniques widely used in safety-critical systems are being adapted to support the analysis of secure systems. Perhaps the most relevant system safety techique for Intrusion Detection Systems is hazard analysis.

  6. Safety in the Chemical Laboratory: Flood Control.

    Science.gov (United States)

    Pollard, Bruce D.

    1983-01-01

    Describes events leading to a flood in the Wehr Chemistry Laboratory at Marquette University, discussing steps taken to minimize damage upon discovery. Analyzes the problem of flooding in the chemical laboratory and outlines seven steps of flood control: prevention; minimization; early detection; stopping the flood; evaluation; clean-up; and…

  7. Chemical safety of food and drinking water

    International Nuclear Information System (INIS)

    Food and drinking water are major sources of human exposure to a large number of chemicals added intentionally for technological reasons or present unintentionally due to contamination. On the other hand, there is a public demand for an essentially risk-free supply of food and drinking water. The concern over the presence of chemicals in the human diet received further emphasis through the development of toxicological and analytical methodology with increased sensitivity over the years. In order to minimize the potential health hazards to the consumers, standards have been established which indicate levels of consumption that are - according to scientific evidence - considered safe and which, consequently, permit control measures to be taken. In this context, public perception of a particular risk, may not always be in line with what might be considered a 'real' risk. Thus, while in the public opinion risk associated with smoking or over-nutrition might be accepted or underestimated, certain food chemical related risks may not be accepted and are sometimes perceived as alarmingly high

  8. Performance indicators for monitoring safety management systems in chemical industry

    Directory of Open Access Journals (Sweden)

    M. Jovašević-Stojanović

    2009-01-01

    Full Text Available The development of the Safety Management System (SMS in chemical industry appears as one of the important requirements introduced by the EU "Seveso II" Directive on the control of major-accident hazards. This paper aims to provide a contribution regarding the SMS structure and the definition of the tools for assessing the effectiveness of this system by means of safety performance indicators. The performance indicators are linked to a reference values or policy targets, illustrating how far the SMS is from the desired level. We developed a system of performance indicators for SMS in chemical industry by using the concept of environmental performance indicators defined in standard ISO 14031. A set of three types of safety system performance indicators was proposed: management performance indicators, operational performance indicators and safety status indicators. These indicators represent the most important factors in the linkage between a possible cause of an accident and its effects.

  9. Chemical Safety Vulnerability Working Group report. Volume 2

    International Nuclear Information System (INIS)

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 2 consists of seven appendices containing the following: Tasking memorandums; Project plan for the CSV Review; Field verification guide for the CSV Review; Field verification report, Lawrence Livermore National Lab.; Field verification report, Oak Ridge Reservation; Field verification report, Savannah River Site; and the Field verification report, Hanford Site

  10. Chemical Safety Vulnerability Working Group report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 2 consists of seven appendices containing the following: Tasking memorandums; Project plan for the CSV Review; Field verification guide for the CSV Review; Field verification report, Lawrence Livermore National Lab.; Field verification report, Oak Ridge Reservation; Field verification report, Savannah River Site; and the Field verification report, Hanford Site.

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

    International Nuclear Information System (INIS)

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

  12. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

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

  13. Chemical process safety management within the Department of Energy

    International Nuclear Information System (INIS)

    Although the Department of Energy (DOE) is not well known for its chemical processing activities, the DOE does have a variety of chemical processes covered under OSHA's Rule for Process Safety Management of Highly Hazardous Chemicals (the PSM Standard). DOE, like industry, is obligated to comply with the PSM Standard. The shift in the mission of DOE away from defense programs toward environmental restoration and waste management has affected these newly forming process safety management programs within DOE. This paper describes the progress made in implementing effective process safety management programs required by the PSM Standard and discusses some of the trends that have supported efforts to reduce chemical process risks within the DOE. In June of 1994, a survey of chemicals exceeding OSHA PSM or EPA Risk Management Program threshold quantities (TQs) at DOE sites found that there were 22 processes that utilized toxic or reactive chemicals over TQs; there were 13 processes involving flammable gases and liquids over TQs; and explosives manufacturing occurred at 4 sites. Examination of the survey results showed that 12 of the 22 processes involving toxic chemicals involved the use of chlorine for water treatment systems. The processes involving flammable gases and liquids were located at the Strategic Petroleum Reserve and Naval petroleum Reserve sites

  14. Relationship of green chemistry and chemical environment safety management

    Institute of Scientific and Technical Information of China (English)

    NieJL; ShenYW

    2002-01-01

    Green chemistry and chemical environmental safety management are the two important techniques and management means to implement sustainable development policy.They are also the two basic tools to carry out headstream depollution and environmental protection.This paper reviewed the principle of green chemistry and main contents of chemical environment safety management from the point of management toxicity,pointed out the same aim of these two techniques and management measures,and described the foreground of those two sustainable development environmental methods in China.

  15. Savannah River Site management response plan for chemical safety vulnerability field assessment. Revision 1

    International Nuclear Information System (INIS)

    As part of the U.S. Department of Energy's (DOE) initiative to identify potential chemical safety vulnerabilities in the DOE complex, the Chemical Safety Vulnerability Core Working Group issued a field verification assessment report. While the report concluded that Savannah River Site (SRS) is moving in a positive direction, the report also identified five chemical safety vulnerabilities with broad programmatic impact that are not easily nor quickly remedied. The May 1994 SRS Management Response Plan addressed the five SRS vulnerabilities identified in the field assessment report. The SRS response plan listed observations supporting the vulnerabilities and any actions taken or planned toward resolution. Many of the observations were resolved by simple explanations, such as the existence of implementation plans for Safety Analysis Report updates. Recognizing that correcting individual observations does not suffice in remedying the vulnerabilities, a task team was assembled to address the broader programmatic issues and to recommend corrective actions

  16. Savannah River Site management response plan for chemical safety vulnerability field assessment. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kahal, E.J.; Murphy, S.L.; Salaymeh, S.R.

    1994-09-01

    As part of the U.S. Department of Energy`s (DOE) initiative to identify potential chemical safety vulnerabilities in the DOE complex, the Chemical Safety Vulnerability Core Working Group issued a field verification assessment report. While the report concluded that Savannah River Site (SRS) is moving in a positive direction, the report also identified five chemical safety vulnerabilities with broad programmatic impact that are not easily nor quickly remedied. The May 1994 SRS Management Response Plan addressed the five SRS vulnerabilities identified in the field assessment report. The SRS response plan listed observations supporting the vulnerabilities and any actions taken or planned toward resolution. Many of the observations were resolved by simple explanations, such as the existence of implementation plans for Safety Analysis Report updates. Recognizing that correcting individual observations does not suffice in remedying the vulnerabilities, a task team was assembled to address the broader programmatic issues and to recommend corrective actions.

  17. Functional Hazard Analysis for Railway Safety

    OpenAIRE

    RAFRAFI,M; El-Koursi, Em

    2007-01-01

    The apportionment of railway safety targets is a key issue to develop a common safety management in the European railway system. In this paper, we develop a generic approach based on the Functional Hazard Analysis (FHA), to analyse the safety of railway systems for a unified European network and to comply with the Common Safety Targets (CSTs) required by the European railway safety directive. We suggest to combine the FHA technique with the functional railway architecture, developed by the AE...

  18. Chemical detection, identification, and analysis system

    International Nuclear Information System (INIS)

    The chemical detection, identification, and analysis system (CDIAS) has three major goals. The first is to display safety information regarding chemical environment before personnel entry. The second is to archive personnel exposure to the environment. Third, the system assists users in identifying the stage of a chemical process in progress and suggests safety precautions associated with that process. In addition to these major goals, the system must be sufficiently compact to provide transportability, and it must be extremely simple to use in order to keep user interaction at a minimum. The system created to meet these goals includes several pieces of hardware and the integration of four software packages. The hardware consists of a low-oxygen, carbon monoxide, explosives, and hydrogen sulfide detector; an ion mobility spectrometer for airborne vapor detection; and a COMPAQ 386/20 portable computer. The software modules are a graphics kernel, an expert system shell, a data-base management system, and an interface management system. A supervisory module developed using the interface management system coordinates the interaction of the other software components. The system determines the safety of the environment using conventional data acquisition and analysis techniques. The low-oxygen, carbon monoxide, hydrogen sulfide, explosives, and vapor detectors are monitored for hazardous levels, and warnings are issued accordingly

  19. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

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

  20. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

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

  1. Technical safety appraisal of the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    On June 27, 1989, Secretary of Energy, Admiral James D. Watkins, US Navy (Retired), announced a 10-point initiative to strengthen environment, safety, and health (ES ampersand H) programs and waste management operations in the Department of Energy (DOE). One of the initiatives involved conducting independent Tiger Team Assessments (TTA) at DOE operating facilities. A TTA of the Idaho National Engineering Laboratory (INEL) was performed during June and July 1991. Technical Safety Appraisals (TSA) were conducted in conjunction with the TTA as its Safety and Health portion. However, because of operational constraints the the Idaho Chemical Processing Plant (ICPP), operated for the DOE by Westinghouse Idaho Nuclear Company, Inc. (WINCO), was not included in the Safety and Health Subteam assessment at that time. This TSA, conducted April 12 - May 8, 1992, was performed by the DOE Office of Performance Assessment to complete the normal scope of the Safety and Health portion of the Tiger Team Assessment of the Idaho National Engineering Laboratory. The purpose of TSAs is to evaluate and strengthen DOE operations by verifying contractor compliance with DOE Orders, to assure that lessons learned from commercial operations are incorporated into facility operations, and to stimulate and encourage pursuit of excellence; thus, the appraisal addresses more issues than would be addressed in a strictly compliance-oriented appraisal. A total of 139 Performance Objectives have been addressed by this appraisal in 19 subject areas. These 19 areas are: organization and administration, quality verification, operations, maintenance, training and certification, auxiliary systems, emergency preparedness, technical support, packaging and transportation, nuclear criticality safety, safety/security interface, experimental activities, site/facility safety review, radiological protection, worker safety and health compliance, personnel protection, fire protection, medical services and natural

  2. Chemical Analysis Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Uses state-of-the-art instrumentation for qualitative and quantitative analysis of organic and inorganic compounds, and biomolecules from gas, liquid, and...

  3. Conservation of Life as a Unifying Theme for Process Safety in Chemical Engineering Education

    Science.gov (United States)

    Klein, James A.; Davis, Richard A.

    2011-01-01

    This paper explores the use of "conservation of life" as a concept and unifying theme for increasing awareness, application, and integration of process safety in chemical engineering education. Students need to think of conservation of mass, conservation of energy, and conservation of life as equally important in engineering design and analysis.…

  4. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  5. Chemical substructure analysis in toxicology

    International Nuclear Information System (INIS)

    A preliminary examination of chemical-substructure analysis (CSA) demonstrates the effective use of the Chemical Abstracts compound connectivity file in conjunction with the bibliographic file for relating chemical structures to biological activity. The importance of considering the role of metabolic intermediates under a variety of conditions is illustrated, suggesting structures that should be examined that may exhibit potential activity. This CSA technique, which utilizes existing large files accessible with online personal computers, is recommended for use as another tool in examining chemicals in drugs. 2 refs., 4 figs

  6. Chemical substructure analysis in toxicology

    Energy Technology Data Exchange (ETDEWEB)

    Beauchamp, R.O. Jr. [Center for Information on Toxicology and Environment, Raleigh, NC (United States)

    1990-12-31

    A preliminary examination of chemical-substructure analysis (CSA) demonstrates the effective use of the Chemical Abstracts compound connectivity file in conjunction with the bibliographic file for relating chemical structures to biological activity. The importance of considering the role of metabolic intermediates under a variety of conditions is illustrated, suggesting structures that should be examined that may exhibit potential activity. This CSA technique, which utilizes existing large files accessible with online personal computers, is recommended for use as another tool in examining chemicals in drugs. 2 refs., 4 figs.

  7. Safety, health and environmental committee (JKSHE): Establishing chemical hazard management

    International Nuclear Information System (INIS)

    Most of the laboratories in Malaysian Nuclear Agency are using chemicals in their research activities. However, it is known that using of chemicals without proper knowledge especially on the material characteristics as well as safe handling procedure may cause great harm to the workers. Therefore, Safety, Health and Environmental Committee (JKSHE) sees the need to establish a good chemical hazard management to ensure that a safe and healthy workplace and environment is provided. One of the elements in chemical hazard management is to carry out Chemical Hazard Risk Assessment (CHRA). The assessment was done so that decision can be made on suitable control measures upon use of such chemicals, such as induction and training courses to be given to the workers and health surveillance activities that may be needed to protect the workers. For this, JKSHE has recommended to conduct CHRA for one of the laboratories at Secondary Standard Dosimetry Laboratory (SSDL) namely Film Dosimeter Processing Room (dark room) as the initial effort towards a better chemical hazard management. This paper presents the case study where CHRA was conducted to identify the chemical hazards at the selected laboratory, the adequacy of existing control measures and finally the recommendation for more effective control measures. (author)

  8. Automation for System Safety Analysis

    Science.gov (United States)

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

    2009-01-01

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

  9. Process Control Systems in the Chemical Industry: Safety vs. Security

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Hahn; Thomas Anderson

    2005-04-01

    Traditionally, the primary focus of the chemical industry has been safety and productivity. However, recent threats to our nation’s critical infrastructure have prompted a tightening of security measures across many different industry sectors. Reducing vulnerabilities of control systems against physical and cyber attack is necessary to ensure the safety, security and effective functioning of these systems. The U.S. Department of Homeland Security has developed a strategy to secure these vulnerabilities. Crucial to this strategy is the Control Systems Security and Test Center (CSSTC) established to test and analyze control systems equipment. In addition, the CSSTC promotes a proactive, collaborative approach to increase industry's awareness of standards, products and processes that can enhance the security of control systems. This paper outlines measures that can be taken to enhance the cybersecurity of process control systems in the chemical sector.

  10. 14 CFR 35.15 - Safety analysis.

    Science.gov (United States)

    2010-01-01

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

  11. 14 CFR 33.75 - Safety analysis.

    Science.gov (United States)

    2010-01-01

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

  12. CHEMICAL PLANT SAFETY AND LOSS PREVENTION (Papers Presented at the International Symposium on Safety Control and Risk Management, SCRM)

    OpenAIRE

    Smith, Robert A.; Michigan, Midland

    1989-01-01

    Increased emphasis on safety and loss prevention over the last 50 years has engrained safety as one of the core values of The Dow Chemical Company. The safety emphasis starts at the very top, with the Environment, Health and Safety Committee of the Board

  13. Regulation of chemical safety at fuel cycle facilities by the United States Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    When the U.S. Nuclear Regulatory Commission (NRC) was established in 1975, its regulations were based on radiation dose limits. Chemical hazards rarely influenced NRC regulations. After the Three Mile Island reactor accident in 1979, the NRC staff was directed to address emergency planning at non-reactor facilities. Several fuel cycle facilities were ordered to submit emergency plans consistent with reactor emergency plans because no other guidance was available. NRC published a notice that it was writing regulations to codify the requirements in the Orders and upgrade the emergency plans to address all hazards, including chemical hazards. The legal authority of NRC to regulate chemical safety was questioned. In 1986, an overfilled uranium hexafluoride cylinder ruptured and killed a worker. The NRC staff was directed to address emergency planning for hazardous chemicals in its regulations. The final rule included a requirement for fuel cycle facilities to certify compliance with legislation requiring local authorities to establish emergency plans for hazardous chemicals. As with emergency planning, NRC's authority to regulate chemical safety during routine operations was limited. NRC established memoranda of understanding (MOUs) with other regulatory agencies to encourage exchange of information between the agencies regarding occupational hazards. In 2000, NRC published new, performance-based, regulations for fuel cycle facilities. The new regulations required an integrated safety analysis (ISA) which used quantitative standards to assess chemical exposures. Some unique chemical exposure cases were addressed while implementing the new regulations. In addition, some gaps remain in the regulation of hazardous chemicals at fuel cycle facilities. The status of ongoing efforts to improve regulation of chemical safety at fuel cycle facilities is discussed. (authors)

  14. NKS/SOS-1 seminar on safety analysis

    International Nuclear Information System (INIS)

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

  15. NKS/SOS-1 seminar on safety analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-01

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

  16. Task D: Hydrogen safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Swain, M.R.; Sievert, B.G. [Univ. of Miami, Coral Gables, FL (United States); Swain, M.N. [Analytical Technologies, Inc., Miami, FL (United States)

    1996-10-01

    This report covers two topics. The first is a review of codes, standards, regulations, recommendations, certifications, and pamphlets which address safety of gaseous fuels. The second is an experimental investigation of hydrogen flame impingement. Four areas of concern in the conversion of natural gas safety publications to hydrogen safety publications are delineated. Two suggested design criteria for hydrogen vehicle fuel systems are proposed. It is concluded from the experimental work that light weight, low cost, firewalls to resist hydrogen flame impingement are feasible.

  17. Guidance for preparation of safety analysis reports

    International Nuclear Information System (INIS)

    Department of Energy (DOE) Order 5480.5, ''Safety of Nuclear Facilities,'' requires the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Operations (ORO) nonreactor facilities and operations. This guide provides a narrative outline of the minimum information needed to prepare safety documentation for ORO moderate or high hazard facilities. Safety documentation is required for new, existing, and modified facilities. The basic purpose of the safety documentation process is to provide assurance that there is no significant increase in risk, as defined in DOE safety policy statements, to people or the environment from operation of ORO facilities

  18. Applications of integrated safety analysis methodology to reload safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Chan Su; Um, Kil Sup [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2011-03-15

    Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the integrated safety analysis methodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, automatic steady-state initialization and safety analysis tool (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants

  19. Applications of integrated safety analysis methodology to reload safety evaluation

    International Nuclear Information System (INIS)

    Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the integrated safety analysis methodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, automatic steady-state initialization and safety analysis tool (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants

  20. Hot Cell Facility (HCF) Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

  1. Hot Cell Facility (HCF) Safety Analysis Report

    International Nuclear Information System (INIS)

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

  2. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

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

  3. Probabilistic Model-Based Safety Analysis

    CERN Document Server

    Güdemann, Matthias; 10.4204/EPTCS.28.8

    2010-01-01

    Model-based safety analysis approaches aim at finding critical failure combinations by analysis of models of the whole system (i.e. software, hardware, failure modes and environment). The advantage of these methods compared to traditional approaches is that the analysis of the whole system gives more precise results. Only few model-based approaches have been applied to answer quantitative questions in safety analysis, often limited to analysis of specific failure propagation models, limited types of failure modes or without system dynamics and behavior, as direct quantitative analysis is uses large amounts of computing resources. New achievements in the domain of (probabilistic) model-checking now allow for overcoming this problem. This paper shows how functional models based on synchronous parallel semantics, which can be used for system design, implementation and qualitative safety analysis, can be directly re-used for (model-based) quantitative safety analysis. Accurate modeling of different types of proba...

  4. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  5. The 'PROCESO' index: a new methodology for the evaluation of operational safety in the chemical industry

    International Nuclear Information System (INIS)

    The acknowledgement of industrial installations as complex systems in the early 1980s outstands as a milestone in the path to operational safety. Process plants are social-technical complex systems of a dynamic nature, whose properties depend not only on their components, but also on the inter-relations among them. A comprehensive assessment of operational safety requires a systemic approach, i.e. an integrated framework that includes all the relevant factors influencing safety. Risk analysis methodologies and safety management systems head the list of methods that point in this direction, but they normally require important plant resources. As a consequence, their use is frequently restricted to especially dangerous processes often driven by compliance with legal requirements. In this work a new safety index for the chemical industry, termed the 'Proceso' Index (standing for the Spanish terms for PROCedure for the Evaluation of Operational Safety), has been developed. PROCESO is based on the principles of systems theory, has a tree-like structure and considers 25 areas to guide the review of plant safety. The method uses indicators whose respective weight values have been obtained via an expert judgement technique. This paper describes the steps followed to develop this new Operational Safety Index, explains its structure and illustrates its application to process plants

  6. Microfabricated Chemical Sensors for Safety and Emission Control Applications

    Science.gov (United States)

    Hunter, G. W.; Neudeck, P. G.; Chen, L.-Y.; Knight, D.; Liu, C. C.; Wu, Q. H.

    1998-01-01

    Chemical sensor technology is being developed for leak detection, emission monitoring, and fire safety applications. The development of these sensors is based on progress in two types of technology: 1) Micromachining and microfabrication (MicroElectroMechanical Systems (MEMS)-based) technology to fabricate miniaturized sensors. 2) The development of high temperature semiconductors, especially silicon carbide. Using these technologies, sensors to measure hydrogen, hydrocarbons, nitrogen oxides, carbon monoxide, oxygen, and carbon dioxide are being developed. A description is given of each sensor type and its present stage of development. It is concluded that microfabricated sensor technology has significant potential for use in a range of aerospace applications.

  7. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

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

  8. Autoclave nuclear criticality safety analysis

    International Nuclear Information System (INIS)

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF6. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF6 inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF6 enriched to 5 percent U235. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF6. 4 refs., 3 figs

  9. Autoclave nuclear criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-31

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

  10. A sequential-move game for enhancing safety and security cooperation within chemical clusters

    International Nuclear Information System (INIS)

    The present paper provides a game theoretic analysis of strategic cooperation on safety and security among chemical companies within a chemical industrial cluster. We suggest a two-stage sequential move game between adjacent chemical plants and the so-called Multi-Plant Council (MPC). The MPC is considered in the game as a leader player who makes the first move, and the individual chemical companies are the followers. The MPC's objective is to achieve full cooperation among players through establishing a subsidy system at minimum expense. The rest of the players rationally react to the subsidies proposed by the MPC and play Nash equilibrium. We show that such a case of conflict between safety and security, and social cooperation, belongs to the 'coordination with assurance' class of games, and we explore the role of cluster governance (fulfilled by the MPC) in achieving a full cooperative outcome in domino effects prevention negotiations. The paper proposes an algorithm that can be used by the MPC to develop the subsidy system. Furthermore, a stepwise plan to improve cross-company safety and security management in a chemical industrial cluster is suggested and an illustrative example is provided.

  11. A sequential-move game for enhancing safety and security cooperation within chemical clusters.

    Science.gov (United States)

    Pavlova, Yulia; Reniers, Genserik

    2011-02-15

    The present paper provides a game theoretic analysis of strategic cooperation on safety and security among chemical companies within a chemical industrial cluster. We suggest a two-stage sequential move game between adjacent chemical plants and the so-called Multi-Plant Council (MPC). The MPC is considered in the game as a leader player who makes the first move, and the individual chemical companies are the followers. The MPC's objective is to achieve full cooperation among players through establishing a subsidy system at minimum expense. The rest of the players rationally react to the subsidies proposed by the MPC and play Nash equilibrium. We show that such a case of conflict between safety and security, and social cooperation, belongs to the 'coordination with assurance' class of games, and we explore the role of cluster governance (fulfilled by the MPC) in achieving a full cooperative outcome in domino effects prevention negotiations. The paper proposes an algorithm that can be used by the MPC to develop the subsidy system. Furthermore, a stepwise plan to improve cross-company safety and security management in a chemical industrial cluster is suggested and an illustrative example is provided. PMID:21146296

  12. Safety analysis of the VLJ repository

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    EVANS, C B

    2004-12-21

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

  14. Economic analysis of safety risks in construction

    OpenAIRE

    Teresa Bourbon; Fernando Santos; Alfredo Soeiro

    2007-01-01

    The objective of this study revolves around the analysis of the safety risks involved with one construction project, and the respective economic effects of risk prevention and safety management. As a result of the co-ordination of systems, and harmonising of work between the Project Leader, Safety Co-ordinator and Contractor, an adequate strategy was developed for the safety of the project Escola de Ciências da Saúde da Universidade do Minho. The risk evaluation is carried out in simulated fo...

  15. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  16. Development of Safety Analysis Technology for LMR

    International Nuclear Information System (INIS)

    In the safety analysis code system development area, the development of an analysis code for a flow blockage could be brought to completion throughout an integrated validation of MATRA-LMR-FB. The safety analysis code of SSC-K has been evolved by building detailed reactivity models and a core 3 dimensional T/H model into it, and developing its window version. A basic analysis module for SFR features also have been developed incorporating a numerical method, best estimated correlations, and a code structure module. For the analysis of the HCDA initiating phase, a sodium boiling model to be linked to SSC-K and a fuel transient performance/cladding failure model have been developed with a state-of-the-art study on the molten fuel movement models. Besides, scoping analysis models for the post-accident heat removal phase have been developed as well. In safety analysis area, the safety criteria for the KALIMER-600 have been set up, and an internal flow channel blockage and local faults have been analyzed for the assembly safety evaluation, while key safety concepts of the KALIMER-600 has been investigated getting through the analyses of ATWS as well as design basis accidents like TOP and LOF, from which the inherent safety due to a core reactivity feedback has been assessed. The HCDA analysis for the initiating phase and an estimation of the core energy release, subsequently, have been followed with setup of the safety criteria as well as T/H analysis for the core catcher. The thermal-hydraulic behaviors, and released radioactivity sources and dose rates in the containment have been analyzed for its performance evaluation in this area. The display of a data base for research products on the KALIMER Website and the detailed process planning with its status analysis, have become feasible from achievements in the area of the integrated technology development and establishment

  17. Safety analysis in subsurface repositories

    International Nuclear Information System (INIS)

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

  18. Safety analysis for non-power reactors

    International Nuclear Information System (INIS)

    Non-power reactors have been operating in Canada since 1945, with NRU (National Research Universal, 1957) being the oldest operating non-power reactor. Presently, there are five generic 'types' of non-power reactors: NRU, ZED-2, SLOWPOKE, MNR and MAPLE, the latter undergoing commissioning as the MDS Medical Isotope Reactor. These reactors range in thermal power from 200 Watts to more than 100 MW. Other non-power reactors are likely to be built for new applications and to replace older reactors. The uniqueness of each reactor, the wide range of power levels and the evolution of safety philosophy over time have lead to non-uniform practices for safety analysis. This non-uniformity may be a problem for the preparation by the licensee and review by the regulator of the safety analysis report required for licensing of the reactor facility. Clearly, there is no universally applicable practice, while at the same time, expectations for safety analyses have evolved in order to demonstrate higher levels of overall safety. This paper examines a new 'graded approach' to preparing the safety analysis report for reactors of diverse features but with a common standard of safety. It discusses necessary content, methods and the training and qualification of the safety analyst. (author)

  19. Preliminary safety analysis for the Gorleben site

    International Nuclear Information System (INIS)

    A methodological approach for long-term safety analyses was applied in preparing the project ''Preliminary Safety Analysis of the Gorleben Site'' based on the German requirements for final disposal of heat generating radioactive waste. The approach included the description of the geological site and its future evolution coupled with a range of possible waste emplacement scenarios using site specific repository designs. The repository designs were developed to provide operational safety, long-term safety and retrieval / recovering of the waste to comply with the German requirements (BMU 2010). A safety concept for accomplishment of the radiological safety and its assessment was developed. Analysis of the integrity, fluid flows and radiological consequences were done and the performance of the system was assessed. If the present assumptions are confirmed by future research and development, the designed repository system for drift disposal was assessed to be robust. Further optimization of the present drift disposal design is possible in relation to gaseous radionuclide species and in relation to the technology of retrieval for the of borehole disposal design. This site specific safety analysis identified important tasks for research and development. The methodology and safety concept can be applied to other salt rock sites and can be transferred partially to clay sites.

  20. Software safety analysis practice in installation phase

    International Nuclear Information System (INIS)

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

  1. Safety Analysis of an Evolving Software Architecture

    OpenAIRE

    de Lemos, Rogério

    2000-01-01

    The safety analysis of an evolving software system has to consider the impact that changes might have on the software components, and to provide confidence that the risk is acceptable. If the impact of a change is not thoroughly analysed, accidents can occur as a result of faulty interactions between components, for example. However, the process of safety analysis can be enhanced if appropriate abstractions are provided for modelling and analysing software components and their interactions. I...

  2. 29 CFR 1910.119 - Process safety management of highly hazardous chemicals.

    Science.gov (United States)

    2010-07-01

    ...: Material Safety Data Sheets meeting the requirements of 29 CFR 1910.1200(g) may be used to comply with this... 29 Labor 5 2010-07-01 2010-07-01 false Process safety management of highly hazardous chemicals... § 1910.119 Process safety management of highly hazardous chemicals. Purpose. This section...

  3. Release mitigation spray safety systems for chemical demilitarization applications.

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, Jonathan; Tezak, Matthew Stephen; Brockmann, John E.; Servantes, Brandon; Sanchez, Andres L.; Tucker, Mark David; Allen, Ashley N.; Wilson, Mollye C.; Lucero, Daniel A.; Betty, Rita G.

    2010-06-01

    Sandia National Laboratories has conducted proof-of-concept experiments demonstrating effective knockdown and neutralization of aerosolized CBW simulants using charged DF-200 decontaminant sprays. DF-200 is an aqueous decontaminant, developed by Sandia National Laboratories, and procured and fielded by the US Military. Of significance is the potential application of this fundamental technology to numerous applications including mitigation and neutralization of releases arising during chemical demilitarization operations. A release mitigation spray safety system will remove airborne contaminants from an accidental release during operations, to protect personnel and limit contamination. Sandia National Laboratories recently (November, 2008) secured funding from the US Army's Program Manager for Non-Stockpile Chemical Materials Agency (PMNSCMA) to investigate use of mitigation spray systems for chemical demilitarization applications. For non-stockpile processes, mitigation spray systems co-located with the current Explosive Destruction System (EDS) will provide security both as an operational protective measure and in the event of an accidental release. Additionally, 'tented' mitigation spray systems for native or foreign remediation and recovery operations will contain accidental releases arising from removal of underground, unstable CBW munitions. A mitigation spray system for highly controlled stockpile operations will provide defense from accidental spills or leaks during routine procedures.

  4. 10 CFR 70.62 - Safety program and integrated safety analysis.

    Science.gov (United States)

    2010-01-01

    ... this safety program; namely, process safety information, integrated safety analysis, and management... conclusion of each failure investigation of an item relied on for safety or management measure. (b) Process safety information. Each licensee or applicant shall maintain process safety information to enable...

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

    International Nuclear Information System (INIS)

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

  6. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

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

  7. Probabilistic safety analysis using microcomputer

    International Nuclear Information System (INIS)

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

  8. Safety Issues of HG and PB as IFE Target Materials: Radiological Versus Chemical Toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Latkowski, J F; Cadwallader, L C; Moir, R W; Rio, G. D; Sanz, J

    2002-11-11

    We have performed a safety assessment of mercury and lead as possible hohlraum materials for Inertial Fusion Energy (IFE) targets, including for the first time a comparative analysis of the radiological and toxicological consequences of an accidental release. In order to calculate accident doses to the public, we have distinguished between accidents at the target fabrication facility and accidents at other areas of the power plant. Regarding the chemical toxicity assessment, we have used the USDOE regulations to determine the maximum allowable release in order to protect the public from adverse health effects. Opposite to common belief, it has been found that the chemical safety requirements for these materials appear to be more stringent than the concentrations that would result in an acceptable radiological dose.

  9. Safety evaluation of chemical mixtures and combinations of chemical and non-chemical stressors.

    Science.gov (United States)

    Jonker, D; Freidig, A P; Groten, J P; de Hollander, A E M; Stierum, R H; Woutersen, R A; Feron, V J

    2004-01-01

    Recent developments in hazard identification and risk assessment of chemical mixtures are reviewed. Empirical, descriptive approaches to study and characterize the toxicity of mixtures have dominated during the past two decades, but an increasing number of mechanistic approaches have made their entry into mixture toxicology. A series of empirical studies with simple chemical mixtures in rats is described in some detail because of the important lessons from this work. The development of regulatory guidelines for the toxicological evaluation of chemical mixtures is discussed briefly. Current issues in mixture toxicology include the adverse health effects of ambient air pollution; the application of such modern, sophisticated methodologies as genomics, bioinformatics, and physiologically based pharmacokinetic modeling; and databases for mixture toxicity. Finally, the state of the art of our knowledge on the potential adverse health effects of combined exposures to chemicals and non-chemical stressors (noise, heat/cold, microorganisms, immobilization, restraint, or transportation), research initiatives in these fields, and the development of an indicator for the cumulative health impact of multiple environmental exposures are discussed. PMID:15329008

  10. Hydrogen Safety Project: Chemical analysis support task

    International Nuclear Information System (INIS)

    Core samples taken from tank 101-SY at Hanford during ''window E'' were analyzed for organic and radiochemical constituents by staff of the Analytical Chemistry Laboratory at Pacific Northwest Laboratory. Westinghouse Hanford company submitted these samples to the laboratory

  11. IMPLEMENTATION OF A SAFETY PROGRAM FOR THE WORK ACCIDENTS’ CONTROL. A CASE STUDY IN THE CHEMICAL INDUSTRY

    Directory of Open Access Journals (Sweden)

    Edison Cesar de Faria Nogueira

    2015-03-01

    Full Text Available This article presents a case study related to the implementation of a Work Safety Program in a chemical industry, based on the Process Safety Program, PSP, of a huge energy company. The research was applied, exploratory, qualitative and with and data collection method through documentary and bibliographical research. There will be presented the main practices adopted in order to make the Safety Program a reality inside a chemical industry, its results and contributions for its better development. This paper proposes the implementation of a Safety Program must be preceded by a diagnosis of occupational safety and health management system and with constant critical analysis in order to make the necessary adjustments.

  12. From Safety Analysis to Formal Specification

    DEFF Research Database (Denmark)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

  14. Chemical analysis by nuclear techniques

    International Nuclear Information System (INIS)

    This state art report consists of four parts, production of micro-particles, analysis of boron, alpha tracking method and development of neutron induced prompt gamma ray spectroscopy (NIPS) system. The various methods for the production of micro-paticles such as mechanical method, electrolysis method, chemical method, spray method were described in the first part. The second part contains sample treatment, separation and concentration, analytical method, and application of boron analysis. The third part contains characteristics of alpha track, track dectectors, pretreatment of sample, neutron irradiation, etching conditions for various detectors, observation of track on the detector, etc. The last part contains basic theory, neutron source, collimator, neutron shields, calibration of NIPS, and application of NIPS system

  15. System safety analysis of an autonomous mobile robot

    International Nuclear Information System (INIS)

    Analysis of the safety of operating and maintaining the Stored Waste Autonomous Mobile Inspector (SWAMI) II in a hazardous environment at the Fernald Environmental Management Project (FEMP) was completed. The SWAMI II is a version of a commercial robot, the HelpMate trademark robot produced by the Transitions Research Corporation, which is being updated to incorporate the systems required for inspecting mixed toxic chemical and radioactive waste drums at the FEMP. It also has modified obstacle detection and collision avoidance subsystems. The robot will autonomously travel down the aisles in storage warehouses to record images of containers and collect other data which are transmitted to an inspector at a remote computer terminal. A previous study showed the SWAMI II has economic feasibility. The SWAMI II will more accurately locate radioactive contamination than human inspectors. This thesis includes a System Safety Hazard Analysis and a quantitative Fault Tree Analysis (FTA). The objectives of the analyses are to prevent potentially serious events and to derive a comprehensive set of safety requirements from which the safety of the SWAMI II and other autonomous mobile robots can be evaluated. The Computer-Aided Fault Tree Analysis (CAFTA copyright) software is utilized for the FTA. The FTA shows that more than 99% of the safety risk occurs during maintenance, and that when the derived safety requirements are implemented the rate of serious events is reduced to below one event per million operating hours. Training and procedures in SWAMI II operation and maintenance provide an added safety margin. This study will promote the safe use of the SWAMI II and other autonomous mobile robots in the emerging technology of mobile robotic inspection

  16. Thermalhydraulic safety analysis of the Candu reactor

    International Nuclear Information System (INIS)

    The thermalhydraulic analysis requirements for the safety and licensing of the CANDU reactor are outlined. The unique features of the CANDU design are first described, and the specialized analysis requirements for the reactor are identified. Thermalhydraulic codes used to perform the analysis are presented and the experimental test programs used to validate the codes are described. The paper concludes with future plans for the experimental test programs, code development, and code validation. (authors). 11 figs., 1 tab., 19 refs

  17. Failure analysis of safety relief valve vibration

    International Nuclear Information System (INIS)

    Shortly after an outage in May 1986, the main steam safety relief valves at the San Onofre Nuclear Generating Station began exhibiting damaging vibration and noise. A systematic failure investigation program was established to identify the root cause. This program included an array of vibration data collection and analysis, as well as inspection and analysis of the main steam piping and safety relief valves. Based upon this investigation, the root cause was traced to two vertical pipe supports which had lost contact with the ground, probably due to some inelastic movement of the piping during heatup and cooldown. Modification of the pipe support eliminated the excessive vibration

  18. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  19. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SEMMENS, L.S.

    1999-02-24

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

  2. Reliability analysis of software based safety functions

    International Nuclear Information System (INIS)

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

  3. 29 CFR 1926.64 - Process safety management of highly hazardous chemicals.

    Science.gov (United States)

    2010-07-01

    ... materials that could foreseeably occur. Note: Material Safety Data Sheets meeting the requirements of 29 CFR... 29 Labor 8 2010-07-01 2010-07-01 false Process safety management of highly hazardous chemicals... Health and Environmental Controls § 1926.64 Process safety management of highly hazardous...

  4. SYNTHESIS OF SAFETY ANALYSIS AND FIRE HAZARD ANALYSIS METHODOLOGIES

    Energy Technology Data Exchange (ETDEWEB)

    Coutts, D

    2007-04-17

    Successful implementation of both the nuclear safety program and fire protection program is best accomplished using a coordinated process that relies on sound technical approaches. When systematically prepared, the documented safety analysis (DSA) and fire hazard analysis (FHA) can present a consistent technical basis that streamlines implementation. If not coordinated, the DSA and FHA can present inconsistent conclusions, which can create unnecessary confusion and can promulgate a negative safety perception. This paper will compare the scope, purpose, and analysis techniques for DSAs and FHAs. It will also consolidate several lessons-learned papers on this topic, which were prepared in the 1990s.

  5. Relationships between psychological safety climate facets and safety behavior in the rail industry: a dominance analysis.

    Science.gov (United States)

    Morrow, Stephanie L; McGonagle, Alyssa K; Dove-Steinkamp, Megan L; Walker, Curtis T; Marmet, Matthew; Barnes-Farrell, Janet L

    2010-09-01

    The goals of this study were twofold: (1) to confirm a relationship between employee perceptions of psychological safety climate and safety behavior for a sample of workers in the rail industry and (2) to explore the relative strengths of relationships between specific facets of safety climate and safety behavior. Non-management rail maintenance workers employed by a large North American railroad completed a survey (n=421) regarding workplace safety perceptions and behaviors. Three facets of safety climate (management safety, coworker safety, and work-safety tension) were assessed as relating to individual workers' reported safety behavior. All three facets were significantly associated with safety behavior. Dominance analysis was used to assess the relative importance of each facet as related to the outcome, and work-safety tension evidenced the strongest relationship with safety behavior. PMID:20538102

  6. Analysis and study on nuclear safety of Mitsubishi PWR

    International Nuclear Information System (INIS)

    Theme of safety analysis and study are changing to reflect the needs at the time. This paper introduces the overall aspects of transient and accident analysis performed and presents typical researches related to safety analysis for Mitsubishi PWR. (author)

  7. Safety analysis review terms of reference

    Energy Technology Data Exchange (ETDEWEB)

    Hurley, T.

    1981-03-01

    This document has been prepared to suggest procedures and items for consideration in the review of safety analysis prepared on DOE fossil energy conversion and technology development projects. It is not intended to reflect official DOE policy. It does, however, provide a basis for consistency in conducting reviews, especially with regard to interpreting levels of risk. Since many of the persons assigned to review panels are not expected to be safety analysts but specialists in related fields such as industrial hygiene and environmental science, this document is intended to provide general terms of reference to facilitate review procedures.

  8. Hanford safety analysis and risk assessment handbook (SARAH)

    International Nuclear Information System (INIS)

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 1,2, and 3 U.S. Department of Energy (DOE) nuclear facilities. SARAH describes currently acceptable methodology for development of a Documented Safety Analysis (DSA) and derivation of technical safety requirements (TSR) based on 10 CFR 830, ''Nuclear Safety Management,'' Subpart B, ''Safety Basis Requirements,'' and provides data to ensure consistency in approach

  9. Deterministic and probabilistic approach to safety analysis

    International Nuclear Information System (INIS)

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

  10. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

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

  11. System analysis for plant operation and safety

    International Nuclear Information System (INIS)

    In parallel with the established reactor support program utilizing design basis system analysis for licensing applications, NUSCO has a broad program underway utilizing best estimate system analysis in support of safe operation of its nuclear units. The latter analysis application requires the use of codes such as RETRAN, which have proven prediction capabilities under a wide range of physical conditions. The program utilizing best estimate system analysis, to varying degrees, in support of plant operation and safety includes the following areas of application: 1) Operator training. Specific application of system analysis in this support area include: best estimate analysis of FSAR transients, best estimate verification of plant specific simulators, and lessons learned through PRA best estimate analysis. 2) Operator guidance. Specific applications in this support area include: development, verification, and safety evaluations of emergency operator guidelines, and analysis of ambiguous scenarios to determine available fail-safe decisions and reversible actions. 3) Operator performance verification. Specific applications in this support area include: verification analysis of operational transients, and verifications of adequacy of system performance/operator actions. 4) Deterministic analyses for PRA support. 5) Verification and support of startup procedures

  12. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  13. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

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

  14. Evaluation model for safety capacity of chemical industrial park based on acceptable regional risk

    Institute of Scientific and Technical Information of China (English)

    Guohua Chen; Shukun Wang; Xiaoqun Tan

    2015-01-01

    The paper defines the Safety Capacity of Chemical Industrial Park (SCCIP) from the perspective of acceptable regional risk. For the purpose of exploring the evaluation model for the SCCIP, a method based on quantitative risk assessment was adopted for evaluating transport risk and to confirm reasonable safety transport capacity of chemical industrial park, and then by combining with the safety storage capacity, a SCCIP evaluation model was put forward. The SCCIP was decided by the smaller one between the largest safety storage capacity and the maximum safety transport capacity, or else, the regional risk of the park will exceed the acceptable level. The developed method was applied to a chemical industrial park in Guangdong province to obtain the maximum safety transport capacity and the SCCIP. The results can be realized in the regional risk control of the park effectively.

  15. Assessing probability of safety criteria exceeding according to probabilistic safety analysis results

    International Nuclear Information System (INIS)

    The paper considers the general task on checking compliance of probabilistic safety indicators with regulatory criteria. It presents correlations to assess probable exceeding of safety criterion for different laws of distribution of the numerical results of the probabilistic safety analysis (PSA). The paper presents the scale for rationing probability of safety criteria exceeding.

  16. Alcator C-MOD final safety analysis

    International Nuclear Information System (INIS)

    This document is designed to address the safety issues involved with the Alcator C-Mod project. This report will begin with a brief description of the experimental objectives which will be followed by information concerning the site. The Alcator C-Mod experiment is a pulsed fusion experiment in which a plasma formed from small amounts of hydrogen or deuterium gas is confined in a magnetic field for short periods (∼1 s). No radioactive fuels or fissile materials are used in the device, so that no criticality hazard exists and no credible nuclear accident can occur. During deuterium operation, the production of a small number of neutrons from a short pulse could result in a small amount of short- and intermediate-lived radioactive isotopes being produced inside the experimental cell. This report will demonstrate that this does not pose an additional hazard to the general population. The health and safety hazards resulting from Alcator C-Mod occur to the workers on the experiment, each of which is described in its own chapter with the steps taken to minimize the risk to employees. These hazards include fire, chemicals and cryogenics, air quality, electrical, electromagnetic radiation, ionizing radiation, and mechanical and natural phenomena. None of these hazards is unique to the facility, and methods of protection from them are well defined and are discussed in the chapter which describes each hazard. The quality assurance program, critical to ensuring the safety aspects of the program, will also be described

  17. Interactive Chemical Safety for Sustainablity Toxicity Forecaster Dashboard

    Data.gov (United States)

    U.S. Environmental Protection Agency — EPA researchers have been using advances in computational toxicology to address lack of data on the thousands of chemicals. EPA released chemical data on 1,800...

  18. Computational methods for nuclear criticality safety analysis

    International Nuclear Information System (INIS)

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

  19. COLD-SAT feasibility study safety analysis

    Science.gov (United States)

    Mchenry, Steven T.; Yost, James M.

    1991-01-01

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

  20. Coulometry in quantitative chemical analysis and physico-chemical research

    International Nuclear Information System (INIS)

    Electroanalytical methods such as potentiometry, amperometry, coulometry and voltammetry are well established and routinely employed in quantitative chemical analysis as well as in chemical research. Coulometry is one of the most important electroanalytical techniques, which involves change in oxidation state of electro active species by heterogeneous electron transfer. In primary coulometric method, uranium is determined at mercury pool electrode and plutonium at platinum gauze electrode

  1. Computer graphics in reactor safety analysis

    International Nuclear Information System (INIS)

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

  2. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

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

  3. Safety analysis report 306-W Building

    International Nuclear Information System (INIS)

    The west portion of the 306 building (306-W), which is operated by PNL, contains a diversified metalworking facility, the BNW specialty shop that machines U, Th, and other weakly radioactive materials, a ceramics laboratory, SNM storage area, and support laboratories, This report presents a safety analysis of the work performed and of the equipment in 306-W. the analyses cover criticality and radiological accidents as well as industrial accidents that could contribute to a criticality or radiological accident

  4. FIND: Standard Safety Analysis Report (GESSAR-238)

    International Nuclear Information System (INIS)

    This index is presented as a guide to microfiche items in Docket STN-50447, which was assigned to the BWR/6 STANDARD SAFETY ANALYSIS REPORT (GESSAR-238) submitted by General Electric Company, San Jose, California. The report describes and analyzes a standard BWR/6 boiling water reactor with a Mark III containment system designed for initial operation at approximately 3579 MW(t) with a net electrical output of approximately 1220 megawatts

  5. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

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

  6. Safety in the Chemical Laboratory: Safety in the Chemistry Laboratories: A Specific Program.

    Science.gov (United States)

    Corkern, Walter H.; Munchausen, Linda L.

    1983-01-01

    Describes a safety program adopted by Southeastern Louisiana University. Students are given detailed instructions on laboratory safety during the first laboratory period and a test which must be completely correct before they are allowed to return to the laboratory. Test questions, list of safety rules, and a laboratory accident report form are…

  7. Comparative analysis of safety related site characteristics

    International Nuclear Information System (INIS)

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

  8. Comparative analysis of safety related site characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Johan (ed.)

    2010-12-15

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

  9. Safety Management Analysis In Construction Industry

    OpenAIRE

    T. Subramani; R. Lordsonmillar

    2014-01-01

    The Indian society and economy have suffered human and financial losses as a result of the poor safety record in the construction industry. The purpose of this study is to examine safety management in the construction industry. The study will collects data from general contractors, who are involved in major types of construction. Collected data include information regarding organizational safety policy, safety training, safety meetings, safety equipment, safety inspections, sa...

  10. Safety analysis of nuclear fuel transport

    International Nuclear Information System (INIS)

    The thermal and structural analysis methods have been improved their efficiency for safety assessments of nuclear fuel transport casks. The pressure-based coupled method recently incorporated in the FLUENT code has been confirmed that it can greatly reduce the calculation time of long term temperature transient analyses for the cask fireproof tests. The parallel computing technique has been investigated for impact load analyses and it is found that by using 32-cores parallel system, the computing time reduces to around 1/10. The pressure-based coupled method and the parallel computing technique will be applied to future expected cross-check analyses and contribute to enhance the quality of the safety evaluation by increasing the number of examination cases. (author)

  11. Methods development for criticality safety analysis

    International Nuclear Information System (INIS)

    A status review on the work at Oak Ridge to develop improved methods for performing multigroup, discrete-ordinates, and Monte Carlo criticality safety analyses is presented. In the area of multigroup cross section preparation this work entails the testing of ENDE/B-IV based and other cross-section libraries in the SCALE system, the development of improved cross-section processing methods for the AMPX system, and the generation of an ENDF/B-V based library. In the area of systems analysis this work entails improvements to the one-dimensional discrete-ordinates code XSDRNPM-S, the testing of the combinatorial geometry version of KENO, KENO-IV/CG, and development of an advanced version of KENO, KENO-V. Also presented is a brief review of the existing criticality safety analytical sequences in the SCALE system, CSAS1 and CSAS2, and the development of the advanced analytical sequences CSAS3 and CSAS4

  12. Coupled seismic analysis of nuclear safety systems

    International Nuclear Information System (INIS)

    Seismic responses of structural systems obtained on the basis of coupled analysis (selected equipment modelled along with the civil structures) results in lower responses and economical designs when compared with uncoupled analysis. For Nuclear Safety Related Structures, from considerations of limiting problem size for analysis and also to reduce modelling efforts, it is necessary to select which equipment needs to be modelled with its supports so as to adequately obtain the response of the structural system with interaction of such equipment. Coupled analysis of a primary structure and secondary system is necessary when the effects of interaction between them are significant. This paper attempts to study the structural response of Reactor Building structures of PHWR as well as PFBR to arrive at specific conclusions with respect to effect of coupling of secondary systems. The paper presents an approach followed to evolve a rational basis for inclusion or non-inclusion of such equipment in the coupled model of the primary system. (author)

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

    International Nuclear Information System (INIS)

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

  14. Management response plan for the Chemical Safety Vulnerability Working Group report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 146 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains a discussion of the chemical safety improvements planned or already underway at DOE sites to correct facility or site-specific vulnerabilities. The main part of the report is a discussion of each of the programmatic deficiencies; a description of the tasks to be accomplished; the specific actions to be taken; and the organizational responsibilities for implementation.

  15. Management response plan for the Chemical Safety Vulnerability Working Group report. Volume 1

    International Nuclear Information System (INIS)

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 146 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains a discussion of the chemical safety improvements planned or already underway at DOE sites to correct facility or site-specific vulnerabilities. The main part of the report is a discussion of each of the programmatic deficiencies; a description of the tasks to be accomplished; the specific actions to be taken; and the organizational responsibilities for implementation

  16. The LaSalle probabilistic safety analysis

    International Nuclear Information System (INIS)

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

  17. ESSAA: Embedded system safety analysis assistant

    Science.gov (United States)

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

    1987-01-01

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

  18. 324 building safety analysis report supplement

    International Nuclear Information System (INIS)

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

  19. Rankine bottoming cycle safety analysis. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lewandowski, G.A.

    1980-02-01

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

  20. Safety strategy and safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

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

  1. Focus on safety : a comparative analysis of pipeline safety performance

    International Nuclear Information System (INIS)

    Canada's National Energy Board (NEB) regulates the design, construction, operation and abandonment of interprovincial and international pipelines within Canada. This publication provides a review of the safety performance of the NEB-regulated oil and gas pipeline industry. It is based on data received through incident reporting under the 1999 Onshore Pipeline Regulations and Safety Performance Indicators. In an effort to assess the performance of NEB-regulated pipelines relative to pipelines constructed under other organizations, a comparison is provided between the performance of NEB-regulated pipeline companies and those of external reference organizations. Details on safety performance from January 1, 2000 to December 31, 2001 were presented. The 6 key indicators that provide comprehensive measures of safety performance for pipeline companies include: fatalities; ruptures; injury frequencies; liquid hydrocarbon releases; gas releases; and, damage prevention. This report reveals that safety has improved in some areas, such as damage due to excavations, but the Board is still concerned about an increase in the rate of contractor injury. However, the Board also acknowledges that accurate projections cannot be made based on 2 years of data. The safety performance of NEB-regulated pipelines is consistent with reference organizations with some exceptions. There was a significant difference in data between 2000 and 2001, some of which is due to the completion of major construction projects. There were no fatalities recorded by NEB-regulated pipeline companies. The number of ruptures increased from 1 in 2000 to 2 in 2001, with the leading cause being pipeline corrosion and the second most common cause being operational errors. The number of spills decreased from 265 in 2000 to 55 in 2001. The high number of spills in 2000 was due, in part, to the high level of construction activity. refs., 6 tabs., 14 figs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. 242-A evaporator safety analysis report

    International Nuclear Information System (INIS)

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

  5. Multilevel analysis in road safety research.

    Science.gov (United States)

    Dupont, Emmanuelle; Papadimitriou, Eleonora; Martensen, Heike; Yannis, George

    2013-11-01

    Hierarchical structures in road safety data are receiving increasing attention in the literature and multilevel (ML) models are proposed for appropriately handling the resulting dependences among the observations. However, so far no empirical synthesis exists of the actual added value of ML modelling techniques as compared to other modelling approaches. This paper summarizes the statistical and conceptual background and motivations for multilevel analyses in road safety research. It then provides a review of several ML analyses applied to aggregate and disaggregate (accident) data. In each case, the relevance of ML modelling techniques is assessed by examining whether ML model formulations (i) allow improving the fit of the model to the data, (ii) allow identifying and explaining random variation at specific levels of the hierarchy considered, and (iii) yield different (more correct) conclusions than single-level model formulations with respect to the significance of the parameter estimates. The evidence reviewed offers different conclusions depending on whether the analysis concerns aggregate data or disaggregate data. In the first case, the application of ML analysis techniques appears straightforward and relevant. The studies based on disaggregate accident data, on the other hand, offer mixed findings: computational problems can be encountered, and ML applications are not systematically necessary. The general recommendation concerning disaggregate accident data is to proceed to a preliminary investigation of the necessity of ML analyses and of the additional information to be expected from their application. PMID:23769622

  6. 242-A evaporator safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    CAMPBELL, T.A.

    1999-05-17

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

  7. Tuning Up BIM for Safety Analysis

    OpenAIRE

    Taiebat, Mojtaba

    2011-01-01

    The construction industry is on the top list of hazardous industries. This justifies the importance of safety research in this industry. Review of the literature identified â fallsâ as the top mortality source in the construction industry. Therefore, this research focuses on falls from heights. Conventional safety practices have held designers responsible for safety of the end-users, and considered constructors responsible for the safety of construction workers. Design for Safety â a...

  8. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

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

  9. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  10. Quantitative Safety and Security Analysis from a Communication Perspective

    OpenAIRE

    Boris Malinowsky; Hans-Peter Schwefel; Oliver Jung

    2015-01-01

    This paper introduces and exemplifies a trade-off analysis of safety and security properties in distributed systems. The aim is to support analysis for real-time communication and authentication building blocks in a wireless communication scenario. By embedding an authentication scheme into a real-time communication protocol for safety-critical scenarios, we can rely on the protocol’s individual safety and security properties. The resulting communication protocol satisfies selected safety and...

  11. Tolerance and safety of superficial chemical peeling with salicylic acid in various facial dermatoses

    OpenAIRE

    Iqbal Zafar; Rahman Simeen; Bari Arfan

    2005-01-01

    BACKGROUND: Chemical peeling is a skin-wounding procedure that may have some potentially undesirable side-effects. AIMS: The present study is directed towards safety concerns associated with superficial chemical peeling with salicylic acid in various facial dermatoses. METHODS: The study was a non-comparative and a prospective one. Two hundred and sixty-eight patients of either sex, aged between 10 to 60 years, undergoing superficial chemical peeling for various facial dermatoses (melasma, ac...

  12. Chemical analysis of water in hydrogeology

    International Nuclear Information System (INIS)

    The aim of the monograph is to give complete information on the chemical analysis of water hydrogeology not only for the students program of Geology study (Bachelor degree study), Engineering Geology and Hydrogeology (Master's degree study) and Engineering Geology (doctoral level study), but also for students from other colleges and universities schools in Slovakia, as well as in the Czech Republic, dealing with the chemical composition of water and its quality, from different perspectives. The benefit would be for professionals with hydrogeological, water and environmental practices, who can find there all the necessary information about proper water sampling, the units used in the chemical analysis of water, expressing the proper chemical composition of water in its various parameters through classification of chemical composition of the water up to the basic features of physical chemistry at thermodynamic calculations and hydrogeochemical modelling.

  13. Systematic safety analysis of old nuclear power plants

    International Nuclear Information System (INIS)

    A program of systematic safety analysis of old nuclear power plants has been engaged by French safety authorities. Beyond the reshaping of safety documents (safety reports, general rules of operation, incidental and accidental procedures, internal emergency plan and manual of quality organization), this examination consisted of an analysis of the operation experience of circuits frequently actuated and a systematic analysis of safety circuits. This paper is based on the presentation of the exercise carried out at the Ardennes nuclear power plant operating for 15 years. This paper reviews also the main studies and modifications engaged on this power plant

  14. Safety Analysis of Soybean Processing for Advanced Life Support

    Science.gov (United States)

    Hentges, Dawn L.

    1999-01-01

    Soybeans (cv. Hoyt) is one of the crops planned for food production within the Advanced Life Support System Integration Testbed (ALSSIT), a proposed habitat simulation for long duration lunar/Mars missions. Soybeans may be processed into a variety of food products, including soymilk, tofu, and tempeh. Due to the closed environmental system and importance of crew health maintenance, food safety is a primary concern on long duration space missions. Identification of the food safety hazards and critical control points associated with the closed ALSSIT system is essential for the development of safe food processing techniques and equipment. A Hazard Analysis Critical Control Point (HACCP) model was developed to reflect proposed production and processing protocols for ALSSIT soybeans. Soybean processing was placed in the type III risk category. During the processing of ALSSIT-grown soybeans, critical control points were identified to control microbiological hazards, particularly mycotoxins, and chemical hazards from antinutrients. Critical limits were suggested at each CCP. Food safety recommendations regarding the hazards and risks associated with growing, harvesting, and processing soybeans; biomass management; and use of multifunctional equipment were made in consideration of the limitations and restraints of the closed ALSSIT.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Analysis of safety culture at Rovno NPP

    International Nuclear Information System (INIS)

    The main concepts of safety culture which relate to safety increase in reactor unit operation, their reliable work, high qualification of personnel and personal responsibility of operators are developed. They will be introduced at the Rovno NPP

  18. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

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

  19. Synthesized safety analysis of fusion system

    International Nuclear Information System (INIS)

    General Methodology of Safety Analysis and Evaluation for Fusion Energy System (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) interim design in the Engineering Design Activities (EDA) stage to identify the candidates of the Design Basis Events (DBEs) stage. These DBEs were compared with those of the ITER design in the Conceptual Design Activities (EDA). As a result, 18 candidates of DBEs were selected for EDA interim design in comparison with 25 DBE candidates for the CDA design. The DBE candidates related to the fuel area were categorized in higher event categories than those of the CDA design due to the increase of the mobile tritium-contained in some components. It is important to reduce the inventory of the tritium absorbed in the dust in the vacuum area as well as in the CDA design. Measures were recommended to reduce the mobile tritium dissolved in the coolant in the single loop due to the increase of this estimated inventory. (author)

  20. Modeling Controller Tasks for Safety Analysis

    Science.gov (United States)

    Brown, Molly; Leveson, Nancy G.

    1998-01-01

    As control systems become more complex, the use of automated control has increased. At the same time, the role of the human operator has changed from primary system controller to supervisor or monitor. Safe design of the human computer interaction becomes more difficult. In this paper, we present a visual task modeling language that can be used by system designers to model human-computer interactions. The visual models can be translated into SpecTRM-RL, a blackbox specification language for modeling the automated portion of the control system. The SpecTRM-RL suite of analysis tools allow the designer to perform formal and informal safety analyses on the task model in isolation or integrated with the rest of the modeled system.

  1. LOFT blowdown experiment safety analysis methodology

    International Nuclear Information System (INIS)

    An unprecedented blowdown experiment safety analysis (ESA) has been performed for the first two scheduled nuclear experiments in the Loss-of-Fluid Test (LOFT) facility. The ESA methodology is a unique approach needed to estimate conservatively the maximum consequences that will occur during an experiment. Through use of this information an acceptable risk in terms of adequate protection of the facility, personnel, and general public can be balanced with the requirements of the experiment program objectives. As an example, one of the LOFT program objectives is to evaluate the performance and effectiveness of emergency core cooling systems (ECCS) while relying on the same ECCSs (and backup ECCSs) to effectively perform as plant protection systems (PPS). The purpose of this paper is to present the LOFT blowdown ESA methodology

  2. Safety analysis of surface haulage accidents

    Energy Technology Data Exchange (ETDEWEB)

    Randolph, R.F.; Boldt, C.M.K.

    1996-12-31

    Research on improving haulage truck safety, started by the U.S. Bureau of Mines, is being continued by its successors. This paper reports the orientation of the renewed research efforts, beginning with an update on accident data analysis, the role of multiple causes in these accidents, and the search for practical methods for addressing the most important causes. Fatal haulage accidents most often involve loss of control or collisions caused by a variety of factors. Lost-time injuries most often involve sprains or strains to the back or multiple body areas, which can often be attributed to rough roads and the shocks of loading and unloading. Research to reduce these accidents includes improved warning systems, shock isolation for drivers, encouraging seatbelt usage, and general improvements to system and task design.

  3. Safety evaluation of chemically modified beta-lactoglobulin administered intravaginally.

    Science.gov (United States)

    Guo, Xuetao; Qiu, Lixia; Wang, Yonghong; Wang, Yue; Meng, Yuanguang; Zhu, Yun; Lu, Lu; Jiang, Shibo

    2016-06-01

    Currently, there is no specific antiviral therapy for treatment of HPV infection. Jiang and colleagues previously reported that anhydride-modified proteins have inhibitory activities against multiple viruses including HPV. Here, we evaluated the safety of 3-hydroxyphthalic anhydride-modified bovine beta-lactoglobulin, designated JB01, vaginally applied in women infected by high-risk HPV. After the vaginal application of JB01 in 38 women for 3 months, no serious adverse events were reported, and normalization of the vaginal micro-environment has been observed. It can be concluded that JB01-BD is safe for vaginal use in HPV-infected women, suggesting its potential application for the treatment of HPV infection. J. Med. Virol. 88:1098-1101, 2016. © 2015 Wiley Periodicals, Inc. PMID:26629967

  4. 77 FR 66638 - The Standard on Process Safety Management of Highly Hazardous Chemicals; Extension of the Office...

    Science.gov (United States)

    2012-11-06

    ... Occupational Safety and Health Administration The Standard on Process Safety Management of Highly Hazardous... the Standard on Process Safety Management of Highly Hazardous Chemicals. DATES: Comments must be... elements of the standard; completing a compilation of written process safety information; performing...

  5. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. 40 CFR 761.253 - Chemical analysis.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Chemical analysis. 761.253 Section 761.253 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) TOXIC SUBSTANCES CONTROL ACT... analysis. (a) Extract PCBs from the standard wipe sample collection medium and clean-up the extracted...

  8. Practicing chemical process safety: a look at the layers of protection

    International Nuclear Information System (INIS)

    This presentation will review a few public perceptions of safety in chemical plants and refineries, and will compare these plant workplace risks to some of the more traditional occupations. The central theme of this paper is to provide a 'within-the-fence' view of many of the process safety practices that world class plants perform to pro-actively protect people, property, profits as well as the environment. It behooves each chemical plant and refinery to have their story on an image-rich presentation to stress stewardship and process safety. Such a program can assure the company's employees and help convince the community that many layers of safety protection within our plants are effective, and protect all from harm

  9. Management response plan for the Chemical Safety Vulnerability Working Group report. Volume 2

    International Nuclear Information System (INIS)

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 146 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. To address the facility-specific and site-specific vulnerabilities, responsible DOE and site-contractor line organizations have developed initial site response plans. These plans, presented as Volume 2 of this Management Response Plan, describe the actions needed to mitigate or eliminate the facility- and site-specific vulnerabilities identified by the CSV Working Group field verification teams. Initial site response plans are described for: Brookhaven National Lab., Hanford Site, Idaho National Engineering Lab., Lawrence Livermore National Lab., Los Alamos National Lab., Oak Ridge Reservation, Rocky Flats Plant, Sandia National Laboratories, and Savannah River Site

  10. Management response plan for the Chemical Safety Vulnerability Working Group report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 146 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. To address the facility-specific and site-specific vulnerabilities, responsible DOE and site-contractor line organizations have developed initial site response plans. These plans, presented as Volume 2 of this Management Response Plan, describe the actions needed to mitigate or eliminate the facility- and site-specific vulnerabilities identified by the CSV Working Group field verification teams. Initial site response plans are described for: Brookhaven National Lab., Hanford Site, Idaho National Engineering Lab., Lawrence Livermore National Lab., Los Alamos National Lab., Oak Ridge Reservation, Rocky Flats Plant, Sandia National Laboratories, and Savannah River Site.

  11. Information Services at the Nuclear Safety Analysis Center.

    Science.gov (United States)

    Simard, Ronald

    This paper describes the operations of the Nuclear Safety Analysis Center. Established soon after an accident at the Three Mile Island nuclear power plant near Harrisburg, Pennsylvania, its efforts were initially directed towards a detailed analysis of the accident. Continuing functions include: (1) the analysis of generic nuclear safety issues,…

  12. Regulation and safety implementation of nanotechnology for chemical enterprises in the Central Europe Space

    Science.gov (United States)

    Falk, A.; Hartl, S.; Sinner, F.

    2013-04-01

    As result of the gradually increasing nanotechnology sector there is the necessity of a contemporary analysis of the present regulations used for nanomaterials, to outline the current situation of the nanotechnology sector, to promote international cooperation and research's coordination to overcome disciplinary boundaries, to fill the gap between more and less experienced regions and to turn investments in R&D in industrial innovations. The general objective of the Central Europe project NANOFORCE, which is developed by national and regional chemistry associations and R&D Centres of the Central Europe area, is to foster the innovative nanotechnology-sector networks across Central Europe regions by bringing together public and private organizations to carry out collaborative and interdisciplinary researches on nanomaterials (in the frame of REACH Regulation) and to turn the most promising laboratory results into innovative industrial applications. To build up a legal advisory board for chemical enterprises starting in nanotechnology, a state of the art report on existing safety procedures and nanotech related regulations was produced to give an overview on currently available regulations used by chemical industries and manufacturing companies within the European region to secure their products. The main emphasis was placed on REACH regulation to search for relevant sections concentrating on nanomaterials which are applicable for nanotechnology. In addition, all relevant directives and amendments of REACH were screened with regard to identify gaps where action is still needed and give possible recommendations for the European Commission. Beyond literature research a questionnaire for producers, users, researchers and financiers was developed with the goal to collect information about the nanotechnology sector in the CE region concerning development, financial status, and international cooperation within joint ventures, safety and nanotoxicology.

  13. Regulation and safety implementation of nanotechnology for chemical enterprises in the Central Europe Space

    International Nuclear Information System (INIS)

    As result of the gradually increasing nanotechnology sector there is the necessity of a contemporary analysis of the present regulations used for nanomaterials, to outline the current situation of the nanotechnology sector, to promote international cooperation and research's coordination to overcome disciplinary boundaries, to fill the gap between more and less experienced regions and to turn investments in R and D in industrial innovations. The general objective of the Central Europe project NANOFORCE, which is developed by national and regional chemistry associations and R and D Centres of the Central Europe area, is to foster the innovative nanotechnology-sector networks across Central Europe regions by bringing together public and private organizations to carry out collaborative and interdisciplinary researches on nanomaterials (in the frame of REACH Regulation) and to turn the most promising laboratory results into innovative industrial applications. To build up a legal advisory board for chemical enterprises starting in nanotechnology, a state of the art report on existing safety procedures and nanotech related regulations was produced to give an overview on currently available regulations used by chemical industries and manufacturing companies within the European region to secure their products. The main emphasis was placed on REACH regulation to search for relevant sections concentrating on nanomaterials which are applicable for nanotechnology. In addition, all relevant directives and amendments of REACH were screened with regard to identify gaps where action is still needed and give possible recommendations for the European Commission. Beyond literature research a questionnaire for producers, users, researchers and financiers was developed with the goal to collect information about the nanotechnology sector in the CE region concerning development, financial status, and international cooperation within joint ventures, safety and nanotoxicology.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Safety Analysis for a Radioisotope Stirling Generator

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL) is conducting safety analyses of various lowpower Radioisotope Stirling Generator (RSG) design concepts for the U. S. Department of Energy. These systems are electrical power generators converting thermal energy from plutonium (238Pu) decay to electrical energy via a Stirling cycle generator. The design and function are similar to the RTG (Radioisotope Thermoelectric Generator) used in space missions since the early 1960's, with a more efficient Stirling cycle generator replacing the proven thermoelectric converter. This paper discusses the methods the INL is employing in the safety analysis effort, along with the software tools, lessons learned, and results. The overall goal of our safety analyses is to determine the probability of an accidental plutonium release over the life of the generator. Historical accident rates for various transportation modes were investigated using event tree methods. Source terms were developed for these accidents including primarily impact, fire, and creep rupture. A negative result was defined as rupture of the tantalum alloy containment vessel surrounding the encapsulated plutonia pellet. Damage due to identified impact accidents was evaluated using non-linear finite element software tools. Material models, gathered from a wide variety of sources, included strain-rate and temperature dependencies on yield strength, strain hardening, and rupture. Both individual component and overall system simulation results will be validated by impact testing to be conducted by Los Alamos National Laboratory. Results from deterministic impact, fire, and creep rupture analyses were integrated into the probabilistic (Monte Carlo) risk assessment by correlation functions relating accident parameters to component damage. This approach presented challenges, which are addressed. Other significant issues include limitations of reliable material data at high temperatures and strain rates and development of a technique to

  17. Criticality safety analysis for mockup facility

    International Nuclear Information System (INIS)

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum Keff is 0.28356 well below than the critical limit, Keff=0.95 at normal condition. In a hypothetical accidental condition, the maximum Keff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. Keff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the Keff increases as the water volume ratio increases. It is also revealed that the Keff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum Keff value is 0.93960 lower than the subcritical limit

  18. Safety relief valve alternate analysis method

    International Nuclear Information System (INIS)

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

  19. Biosensors for functional food safety and analysis.

    Science.gov (United States)

    Lavecchia, Teresa; Tibuzzi, Arianna; Giardi, Maria Teresa

    2010-01-01

    The importance of safety and functionality analysis of foodstuffs and raw materials is supported by national legislations and European Union (EU) directives concerning not only the amount of residues of pollutants and pathogens but also the activity and content of food additives and the health claims stated on their labels. In addition, consumers' awareness of the impact of functional foods' on their well-being and their desire for daily healthcare without the intake pharmaceuticals has immensely in recent years. Within this picture, the availability of fast, reliable, low cost control systems to measure the content and the quality of food additives and nutrients with health claims becomes mandatory, to be used by producers, consumers and the governmental bodies in charge of the legal supervision of such matters. This review aims at describing the most important methods and tools used for food analysis, starting with the classical methods (e.g., gas-chromatography GC, high performance liquid chromatography HPLC) and moving to the use of biosensors-novel biological material-based equipments. Four types of bio-sensors, among others, the novel photosynthetic proteins-based devices which are more promising and common in food analysis applications, are reviewed. A particular highlight on biosensors for the emerging market of functional foods is given and the most widely applied functional components are reviewed with a comprehensive analysis of papers published in the last three years; this report discusses recent trends for sensitive, fast, repeatable and cheap measurements, focused on the detection of vitamins, folate (folic acid), zinc (Zn), iron (Fe), calcium (Ca), fatty acids (in particular Omega 3), phytosterols and phytochemicals. A final market overview emphasizes some practical aspects ofbiosensor applications. PMID:21520718

  20. Impacts on health and safety from transfer/consolidation of nuclear materials and hazardous chemicals

    International Nuclear Information System (INIS)

    Environmental restoration plans at the US Department of Energy (USDOE) Hanford Site calls for transfer/consolidation of ''targets/threats,'' namely nuclear materials and hazardous chemicals. Reductions in the health and safety hazards will depend on the plans implemented. Pacific Northwest Laboratory (PNL) estimated these potential impacts, assuming implementation of the current reference plan and employing ongoing risk and safety analyses. The results indicated the potential for ''significant'' reductions in health and safety hazards in the long term (> 25 years) and a potentially ''noteworthy'' reduction in health hazard in the short term (≤ 25 years)

  1. SCANS, Shipping Cask Design Safety Analysis

    International Nuclear Information System (INIS)

    1 - Description of program or function: SCANS (Shipping Cask Analysis System) is a microcomputer-based system of computer programs and databases for evaluating safety analysis reports on spent fuel shipping casks. SCANS calculates the global response to impact loads, pressure loads, and thermal conditions, providing reviewers with an independent check on analyses submitted by licensees. Analysis options are based on regulatory cases described in the Code of Federal Regulation (1983) and Regulatory Guides published by the NRC in 1977 and 1978. The system is composed of a series of menus and input entry cask analysis, and output display programs. An analysis is performed by preparing the necessary input data and then selecting the appropriate analysis: impact, thermal (heat transfer), thermally- induced stress, or pressure-induced stress. All data are entered through input screens with descriptive data requests, and, where possible, default values are provided. Output (i.e., impact force, moment and shear time histories; impact animation; thermal/stress geometry and thermal/stress element outlines; temperature distributions as iso-contours or profiles; and temperature time histories) is displayed graphically and can also be printed. 2 - Method of solution: Impact analyses use a one-dimensional dynamic beam model. Each node in the beam model has two translational and one rotational degrees of freedom. The impact code uses an explicit time-history integration scheme in which equilibrium is formulated in terms of the global external forces and internal force resultants. This formulation allows the code to track large rigid- body motion. Thus, the oblique impact problem can be calculated from initial impact through essentially rigid-body rotation to secondary impact. Lateral pressure due to lead-slump can also be calculated. Appropriate two-dimensional finite-element meshes are automatically generated for thermal, thermal-stress, and pressure- stress analyses, based on

  2. Chemical analysis of high purity graphite

    International Nuclear Information System (INIS)

    The Sub-Committee on Chemical Analysis of Graphite was organized in April 1989, under the Committee on Chemical Analysis of Nuclear Fuels and Reactor Materials, JAERI. The Sub-Committee carried out collaborative analyses among eleven participating laboratories for the certification of the Certified Reference Materials (CRMs), JAERI-G5 and G6, after developing and evaluating analytical methods during the period of September 1989 to March 1992. The certified values were given for ash, boron and silicon in the CRM based on the collaborative analysis. The values for ten elements (Al, Ca, Cr, Fe, Mg, Mo, Ni, Sr, Ti, V) were not certified, but given for information. Preparation, homogeneity testing and chemical analyses for certification of reference materials were described in this paper. (author) 52 refs

  3. Spectroscopic chemical analysis methods and apparatus

    Science.gov (United States)

    Hug, William F. (Inventor); Reid, Ray D. (Inventor); Bhartia, Rohit (Inventor)

    2013-01-01

    Spectroscopic chemical analysis methods and apparatus are disclosed which employ deep ultraviolet (e.g. in the 200 nm to 300 nm spectral range) electron beam pumped wide bandgap semiconductor lasers, incoherent wide bandgap semiconductor light emitting devices, and hollow cathode metal ion lasers to perform non-contact, non-invasive detection of unknown chemical analytes. These deep ultraviolet sources enable dramatic size, weight and power consumption reductions of chemical analysis instruments. Chemical analysis instruments employed in some embodiments include capillary and gel plane electrophoresis, capillary electrochromatography, high performance liquid chromatography, flow cytometry, flow cells for liquids and aerosols, and surface detection instruments. In some embodiments, Raman spectroscopic detection methods and apparatus use ultra-narrow-band angle tuning filters, acousto-optic tuning filters, and temperature tuned filters to enable ultra-miniature analyzers for chemical identification. In some embodiments Raman analysis is conducted along with photoluminescence spectroscopy (i.e. fluorescence and/or phosphorescence spectroscopy) to provide high levels of sensitivity and specificity in the same instrument.

  4. Intelligent Chemical Sensor Systems for In-space Safety Applications

    Science.gov (United States)

    Hunter, G. W.; Xu, J. C.; Neudeck, P. G.; Makel, D. B.; Ward, B.; Liu, C. C.

    2006-01-01

    Future in-space and lunar operations will require significantly improved monitoring and Integrated System Health Management (ISHM) throughout the mission. In particular, the monitoring of chemical species is an important component of an overall monitoring system for space vehicles and operations. For example, in leak monitoring of propulsion systems during launch, inspace, and on lunar surfaces, detection of low concentrations of hydrogen and other fuels is important to avoid explosive conditions that could harm personnel and damage the vehicle. Dependable vehicle operation also depends on the timely and accurate measurement of these leaks. Thus, the development of a sensor array to determine the concentration of fuels such as hydrogen, hydrocarbons, or hydrazine as well as oxygen is necessary. Work has been on-going to develop an integrated smart leak detection system based on miniaturized sensors to detect hydrogen, hydrocarbons, or hydrazine, and oxygen. The approach is to implement Microelectromechanical Systems (MEMS) based sensors incorporated with signal conditioning electronics, power, data storage, and telemetry enabling intelligent systems. The final sensor system will be self-contained with a surface area comparable to a postage stamp. This paper discusses the development of this "Lick and Stick" leak detection system and it s application to In-Space Transportation and other Exploration applications.

  5. Risk Analysis of Safety-Critical Control Systems

    OpenAIRE

    Karol Rastocny

    2008-01-01

    This paper deals with problems associated with risks analysis of a safety-critical control system. In the paper there are introduced recommendations enabling practical enforceability of risk analysis by the assurance of sufficient objectivity level. In the initial phases of the system lifecycle risk analysis serves for a tolerable hazard rate definition for individual safety relevant functions. In the end of the control system development process the risk analysis (an analysis of failures con...

  6. Service activities of chemical analysis division

    International Nuclear Information System (INIS)

    Progress of the Division during the year of 1988 was described on the service activities for various R and D projects carrying out in the Institute, for the fuel fabrication and conversion plant, and for the post-irradiation examination facility. Relevant analytical methodologies developed for the chemical analysis of an irradiated fuel, safeguards chemical analysis, and pool water monitoring were included such as chromatographic separation of lanthanides, polarographic determination of dissolved oxygen in water, and automation on potentiometric titration of uranium. Some of the laboratory manuals revised were also included in this progress report. (Author)

  7. Safety management - policy, analysis and implementation

    International Nuclear Information System (INIS)

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

  8. Safety Analysis versus Type Inference with Partial Types

    DEFF Research Database (Denmark)

    Schwartzbach, Michael Ignatieff; Palsberg, Jens

    1992-01-01

    Safety analysis is an algorithm for determining if a term in an untyped lambda calculus with constants is safe, i.e., if it does not cause an error during evaluation. This ambition is also shared by algorithms for type inference. Safety analysis and type inference are based on rather different...... perspectives, however. Safety analysis is global in that it can only analyze a complete program. In contrast, type inference is local in that it can analyze pieces of a program in isolation. In this paper we prove that safety analysis is sound, relative to both a strict and a lazy operational semantics. We...... also prove that safety analysis accepts strictly more safe lambda terms than does type inference for simple types. The latter result demonstrates that global program analysis can be more precise than local ones....

  9. Propulsion system safety analysis methodology for commercial transport aircraft

    OpenAIRE

    Knife, S.

    1997-01-01

    Airworthiness certification of commercial transport aircraft requires a safety analysis of the propulsion system to establish that the probability of a failure jeopardising the safety of the aeroplane is acceptably low. The needs and desired features of such a propulsion system safety analysis are discussed, and current techniques and assumptions employed in such analyses are evaluated. It is concluded that current assumptions and techniques are not well suited to predicting...

  10. The Safety "Use Case": Co-Developing Chemical Information Management and Laboratory Safety Skills

    Science.gov (United States)

    Stuart, Ralph B.; McEwen, Leah R.

    2016-01-01

    The 2015 edition of the American Chemical Society's "Guidelines and Evaluation Procedures for Bachelor's Degree Programs" identifies six skill sets that undergraduate chemistry programs should instill in their students. In our roles as support staff for chemistry departments at two different institutions (one a Primarily Undergraduate…

  11. Preventing Agricultural Chemical Exposure: A Safety Program Manual. Participatory Education with Farmworkers in Pesticide Safety.

    Science.gov (United States)

    Wake Forest Univ., Winston-Salem, NC. Dept. of Family and Community Medicine.

    Preventing Agricultural Chemical Exposure among North Carolina Farmworkers (PACE) is a project designed to describe farmworker pesticide exposure and to develop an educational intervention to reduce farmworker pesticide exposure. The PACE project used a community participation framework to ensure that the community played a significant role in…

  12. Safety analysis and evaluation methodology for fusion systems

    International Nuclear Information System (INIS)

    Three issues are critical to the public acceptability of nuclear fusion as an energy system. These are technological feasibility, economic viability and safety. Safety will be especially important when tritium is used as a fuel and the reactor becomes radioactive. As a result of this study a safety analysis and evaluation methodology for fusion systems were developed. In this all the safety-related issues in the fusion system could be integrated and resolved. A general descriptive model, the three principle items to be assured, an approach to safety assurance based on event categorization and the function based safety analysis are all discussed. The usefulness of the methodology was illustrated by the application of the safety evaluation to the R-Tokamak. (author)

  13. Analysis on safety production in coal mines Henan Province

    Institute of Scientific and Technical Information of China (English)

    KONG Liu-an; ZHANG Wen-yong

    2006-01-01

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

  14. Entropy generation reduction through chemical pinch analysis

    International Nuclear Information System (INIS)

    The pinch analysis (PA) concept emerged, late '80s, as one of the methods to address the energy management in the new era of sustainable development. It was derived from combined first and second law analysis, as a technique ensuring a better thermal integration, aiming the minimization of entropy production or, equivalently, exergy destruction by heat exchanger networks (HEN). Although its ascendance from the second law analysis is questionable, the PA reveals as a widespread tool, nowadays, helping in energy savings mostly through a more rational use of utilities. Unfortunately, as principal downside, one should be aware that the global minimum entropy production is seldom attained, since the PA does not tackle the whole plant letting aside the chemical reactors or separation trains. The chemical reactor network (CRN) is responsible for large amounts of entropy generation (exergy losses), mainly due to the combined composition and temperature change. The chemical pinch analysis (CPA) concept focuses on, simultaneously, the entropy generation reduction of both CRN and HEN, while keeping the state and working parameters of the plant in the range of industrial interest. The fundamental idea of CPA is to include the CRN (through the chemical reaction heat developed in reactors) into the HEN and to submit this extended system to the PA. This is accomplished by replacing the chemical reactor with a virtual heat exchanger system producing the same amount of entropy. For an endothermic non-adiabatic chemical reactor, the (stepwise infinitesimal) supply heat δq flows from a source (an external/internal heater) to the stream undergoing the chemical transformation through the reactor, which in turn releases the heat of reaction ΔHR to a virtual cold stream flowing through a virtual cooler. For an exothermic non-adiabatic chemical reactor, the replacement is likewise, but the heat flows oppositely. Thus, in the practice of designing or retrofitting a flowsheet, in order to

  15. Chemical and Plant-Based Insect Repellents: Efficacy, Safety, and Toxicity.

    Science.gov (United States)

    Diaz, James H

    2016-03-01

    Most emerging infectious diseases today are arthropod-borne and cannot be prevented by vaccinations. Because insect repellents offer important topical barriers of personal protection from arthropod-borne infectious diseases, the main objectives of this article were to describe the growing threats to public health from emerging arthropod-borne infectious diseases, to define the differences between insect repellents and insecticides, and to compare the efficacies and toxicities of chemical and plant-derived insect repellents. Internet search engines were queried with key words to identify scientific articles on the efficacy, safety, and toxicity of chemical and plant-derived topical insect repellants and insecticides to meet these objectives. Data sources reviewed included case reports; case series; observational, longitudinal, and surveillance studies; and entomological and toxicological studies. Descriptive analysis of the data sources identified the most effective application of insect repellents as a combination of topical chemical repellents, either N-diethyl-3-methylbenzamide (formerly N, N-diethyl-m-toluamide, or DEET) or picaridin, and permethrin-impregnated or other pyrethroid-impregnated clothing over topically treated skin. The insecticide-treated clothing would provide contact-level insecticidal effects and provide better, longer lasting protection against malaria-transmitting mosquitoes and ticks than topical DEET or picaridin alone. In special cases, where environmental exposures to disease-transmitting ticks, biting midges, sandflies, or blackflies are anticipated, topical insect repellents containing IR3535, picaridin, or oil of lemon eucalyptus (p-menthane-3, 8-diol or PMD) would offer better topical protection than topical DEET alone. PMID:26827259

  16. Software FMEA analysis for safety-related application software

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kozine, I.; Duijm, N.J.; Lauridsen, K. [Risoe National Lab. (Denmark)

    2000-12-01

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

  19. Analysis on Impact Factors of Workplace Safety Investment in Chemical Industry Enterprises%化工企业工作环境安全投资的影响因素分析

    Institute of Scientific and Technical Information of China (English)

    王幼莉; 王江辉

    2011-01-01

    以边际效益理论为基础对企业安全投资过程进行了分析,通过对安全程度水平变化的边际效益和边际成本比较分析,找到最优的安全投入决策点.认为合理地进行工作环境安全投资是提高化工企业安全生产水平和创造经济效益的重要手段.在此基础上具体分析了化工企业对工作环境的安全认识、化工企业员工的安全意识及安全行为的规范和政府规制等因素对化工企业安伞投资效益的影响.%Based on marginal utility theory, enterprises workplace safety investment process was analyzed, and through comparing with marginal benefit and marginal investment of change degree of safety level, the optimal investment point was found. It was believed that the reasonable workplace safety investment would promote safety level and would also be an important method to create economic benefit. Effect of enterprises moral and social accountability, employee's safety consciousness and government safety laws and regulations on safety investment benefits was discussed.

  20. Task Group report to the Assistant Secretary for Environment, Safety and Health on oversight of chemical safety at the Department of Energy. Volume 2, Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    This report presents the results of a preliminary review of chemical safety within the Department of Energy (DOE). The review was conducted by Chemical Safety Oversight Review (CSOR) Teams composed of Office of Environment, Safety and Health (EH) staff members and contractors. The primary objective of the CSOR was to assess, the safety status of DOE chemical operations and identify any significant deficiencies associated with such operations. Significant was defined as any situation posing unacceptable risk, that is, imminent danger or threat to workers, co-located workers, the general public, or the environment, that requires prompt action by EH or the line organizations. A secondary objective of the CSOR was to gather and analyze technical and programmatic information related to chemical safety to be used in conjunction with the longer-range EH Workplace Chemical Accident Risk Review (WCARR) Program. The WCARR Program is part of the ongoing EH oversight of nonnuclear safety at all DOE facilities. `` The program objective is to analyze DOE and industry chemical safety programs and performance and determine the need for additional or improved safety guidance for DOE. During the period June 6, 1992, through July 31, 1992, EH conducted CSORs at five DOE sites. The sites visited were Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 Plant (Y-12), Oak Ridge National Laboratory (ORNL), and Lawrence Livermore National Laboratory (LLNL).

  1. Waste Tank Organic Safety Project: Analysis of liquid samples from Hanford waste tank 241-C-103

    International Nuclear Information System (INIS)

    A suite of physical and chemical analyses has been performed in support of activities directed toward the resolution of an Unreviewed Safety Question concerning the potential for a floating organic layer in Hanford waste tank 241-C-103 to sustain a pool fire. The analysis program was the result of a Data Quality Objectives exercise conducted jointly with staff from Westinghouse Hanford Company and Pacific Northwest Laboratory (PNL). The organic layer has been analyzed for flash point, organic composition including volatile organics, inorganic anions and cations, radionuclides, and other physical and chemical parameters needed for a safety assessment leading to the resolution of the Unreviewed Safety Question. The aqueous layer underlying the floating organic material was also analyzed for inorganic, organic, and radionuclide composition, as well as other physical and chemical properties. This work was conducted to PNL Quality Assurance impact level III standards (Good Laboratory Practices)

  2. Waste Tank Organic Safety Project: Analysis of liquid samples from Hanford waste tank 241-C-103

    Energy Technology Data Exchange (ETDEWEB)

    Pool, K.H.; Bean, R.M.

    1994-03-01

    A suite of physical and chemical analyses has been performed in support of activities directed toward the resolution of an Unreviewed Safety Question concerning the potential for a floating organic layer in Hanford waste tank 241-C-103 to sustain a pool fire. The analysis program was the result of a Data Quality Objectives exercise conducted jointly with staff from Westinghouse Hanford Company and Pacific Northwest Laboratory (PNL). The organic layer has been analyzed for flash point, organic composition including volatile organics, inorganic anions and cations, radionuclides, and other physical and chemical parameters needed for a safety assessment leading to the resolution of the Unreviewed Safety Question. The aqueous layer underlying the floating organic material was also analyzed for inorganic, organic, and radionuclide composition, as well as other physical and chemical properties. This work was conducted to PNL Quality Assurance impact level III standards (Good Laboratory Practices).

  3. Special characteristics of the safety analysis of HWRs

    International Nuclear Information System (INIS)

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

  4. TA-55 Final Safety Analysis Report Comparison Document and DOE Safety Evaluation Report Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Alan Bond

    2001-04-01

    This document provides an overview of changes to the currently approved TA-55 Final Safety Analysis Report (FSAR) that are included in the upgraded FSAR. The DOE Safety Evaluation Report (SER) requirements that are incorporated into the upgraded FSAR are briefly discussed to provide the starting point in the FSAR with respect to the SER requirements.

  5. Safety in the Chemical Laboratory--Chemical Management: A Method for Waste Reduction.

    Science.gov (United States)

    Pine, Stanley H.

    1984-01-01

    Discusses methods for reducing or eliminating waste disposal problems in the chemistry laboratory, considering both economic and environmental aspects of the problems. Proposes inventory control, shared use, solvent recycling, zero effluent, and various means of disposing of chemicals. (JM)

  6. ARIES-ST safety design and analysis

    International Nuclear Information System (INIS)

    Activation and safety analyses were performed for the ARIES-ST design. The ARIES-ST power plant includes a water-cooled copper centerpost. The first wall and shield are made of low activation ferritic steel and cooled with helium. The blanket is also made of ferritic steel with SiC inserts and Li17Pb83 breeder. The divertor plate is made of low activation ferritic steel and uses a tungsten brush as plasma facing component. The power plant has a lifetime of 40 full power years (FPY). However, the centerpost, first wall, inboard shield and blanket were assumed to be replaced every 2.86 FPY. Neutron transmutation of copper resulted in the production of several nickel, cobalt and zinc isotopes. The production of these isotopes resulted in an increase of the time-space average electrical resistivity of the centerpost by about 6% after 2.86 FPY. All of the plant components met the limits for disposal as Class C low-level waste. The off-site doses produced at the onset of an accident are caused by the mobilization of the radioactive inventory present in the plant. Analysis of a loss of coolant accident (LOCA) indicated that the centerpost would reach a maximum temperature of about 1000 deg. C during the accident. In the meantime, the first wall and shield would reach a maximum temperature of about 800 deg. C. A similar divertor LOCA analysis indicated that the front tungsten layer would also reach a maximum temperature of about 800 deg. C. The calculated temperature profiles and available oxidation-driven volatility experimental data were used to calculate the dose at the site boundary under conservative release conditions. The current design produces an effective whole body early dose of 1.88 mSv at the site boundary. In addition, a divertor disruption would only produce an effective whole body early dose of 7.68 μSv at the site boundary

  7. STARS software tool for analysis of reliability and safety

    International Nuclear Information System (INIS)

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

  8. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR

  9. Compositional Safety Analysis using Barrier Certificates

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  10. Materials Safety Data Sheets: the basis for control of toxic chemicals

    Energy Technology Data Exchange (ETDEWEB)

    Ketchen, E.E.; Porter, W.E.

    1979-09-01

    The Material Safety Data Sheets contained in this volume are the basis for the Toxic Chemical Control Program developed by the Industrial Hygiene Department, Health Division, ORNL. The three volumes are the update and expansion of ORNL/TM-5721 and ORNL/TM-5722 Material Safety Data Sheets: The Basis for Control of Toxic Chemicals, Volume I and Volume II. As such, they are a valuable adjunct to the data cards issued with specific chemicals. The chemicals are identified by name, stores catalog number where appropriate, and sequence numbers from the NIOSH Registry of Toxic Effects of Chemical Substances, 1977 Edition, if available. The data sheets were developed and compiled to aid in apprising the employees of hazards peculiar to the handling and/or use of specific toxic chemicals. Space limitation necessitate the use of descriptive medical terms and toxicological abbreviations. A glossary and an abbreviation list were developed to define some of those sometimes unfamiliar terms and abbreviations. The page numbers are keyed to the catalog number in the chemical stores at ORNL.

  11. Thermalhydraulic safety analysis of the Candu reactor

    International Nuclear Information System (INIS)

    The thermalhydraulic analysis requirements for the safety and licensing of the Candu reactor are outlined. The unique features of the Candu design are first described, and the specialized analysis requirements for the reactor are identified. Thermalhydraulic codes used to perform the analysis are presented and the experimental test programs used to validate the codes are described. The paper concludes with future plans for the experimental test programs, code development, and code validation. Future experimental work will largely focus on improving our understanding of the interaction of multiple parallel heated channels under upset conditions. This is, of course, related to the blowdown and refill thermal-hydraulics of full-size flow headers connecting the feeders. An increasing emphasis is currently being placed on refining the instrumentation for our test facilities. An Instrument Development Program has been recently implemented to provide instrumentation not currently available commercially. For example, conductivity probes are being developed to accurately measure the level in the RD-14M headers. As well, neutron scattering tomography is being evaluated to measure the void and void distribution in the heated channels of RD-14M and the CWIT facility. To facilitate easy access to experimental data, a program has been initiated to develop a fully relational data base of thermalhydraulic data obtained from our experimental programs. Software is being developed to display the information in a number of formats, including an 'animated' replay of the experiment, to aid the analyst in interpreting the experimental data. Future code development will also focus on accurately predicting the behaviour of the Candu header/feeder system under loss-of-coolant accident conditions. A multi-dimensional representation of the header will be required. At the same time, the computational efficiency of the codes will have to increase to handle the large number of parallel channels

  12. Gap Analysis Approach for Construction Safety Program Improvement

    Directory of Open Access Journals (Sweden)

    Thanet Aksorn

    2007-06-01

    Full Text Available To improve construction site safety, emphasis has been placed on the implementation of safety programs. In order to successfully gain from safety programs, factors that affect their improvement need to be studied. Sixteen critical success factors of safety programs were identified from safety literature, and these were validated by safety experts. This study was undertaken by surveying 70 respondents from medium- and large-scale construction projects. It explored the importance and the actual status of critical success factors (CSFs. Gap analysis was used to examine the differences between the importance of these CSFs and their actual status. This study found that the most critical problems characterized by the largest gaps were management support, appropriate supervision, sufficient resource allocation, teamwork, and effective enforcement. Raising these priority factors to satisfactory levels would lead to successful safety programs, thereby minimizing accidents.

  13. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

  14. Probabilistic safety analysis and interpretation thereof

    International Nuclear Information System (INIS)

    Increasing use of the instrumentation of PSA is being made in Germany for quantitative technical safety assessment, for example with regard to incidents which must be reported and forwarding of information, especially in the case of modification of nuclear plants. The Commission for Nuclear Reactor Safety recommends regular execution of PSA on a cycle period of ten years. According to the PSA guidance instructions, probabilistic analyses serve for assessing the degree of safety of the entire plant, expressed as the expectation value for the frequency of endangering conditions. The authors describe the method, action sequence and evaluation of the probabilistic safety analyses. The limits of probabilistic safety analyses arise in the practical implementation. Normally the guidance instructions for PSA are confined to the safety systems, so that in practice they are at best suitable for operational optimisation only to a limited extent. The present restriction of the analyses has a similar effect on power output operation of the plant. This seriously degrades the utilitarian value of these analyses for the plant operators. In order to further develop PSA as a supervisory and operational optimisation instrument, both authors consider it to be appropriate to bring together the specific know-how of analysts, manufacturers, plant operators and experts. (orig.)

  15. Probabilistic safety analysis : a new nuclear power plants licensing method

    International Nuclear Information System (INIS)

    After a brief retrospect of the application of Probabilistic Safety Analysis in the nuclear field, the basic differences between the deterministic licensing method, currently in use, and the probabilistic method are explained. Next, the two main proposals (by the AIF and the ACRS) concerning the establishment of the so-called quantitative safety goals (or simply 'safety goals') are separately presented and afterwards compared in their most fundamental aspects. Finally, some recent applications and future possibilities are discussed. (Author)

  16. Safety analysis report for the Waste Storage Facility. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  17. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  18. Analysis of microgravity space experiments Space Shuttle programmatic safety requirements

    Science.gov (United States)

    Terlep, Judith A.

    1996-01-01

    This report documents the results of an analysis of microgravity space experiments space shuttle programmatic safety requirements and recommends the creation of a Safety Compliance Data Package (SCDP) Template for both flight and ground processes. These templates detail the programmatic requirements necessary to produce a complete SCDP. The templates were developed from various NASA centers' requirement documents, previously written guidelines on safety data packages, and from personal experiences. The templates are included in the back as part of this report.

  19. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  20. Chemical abundance analysis of 19 barium stars

    CERN Document Server

    Yang, G C; Spite, M; Chen, Y Q; Zhao, G; Zhang, B; Liu, G Q; Liu, Y J; Liu, N; Deng, L C; Spite, F; Hill, V; Zhang, C X

    2016-01-01

    We aim at deriving accurate atmospheric parameters and chemical abundances of 19 barium (Ba) stars, including both strong and mild Ba stars, based on the high signal-to-noise ratio and high resolution Echelle spectra obtained from the 2.16 m telescope at Xinglong station of National Astronomical Observatories, Chinese Academy of Sciences. The chemical abundances of the sample stars were obtained from an LTE, plane-parallel and line-blanketed atmospheric model by inputting the atmospheric parameters (effective temperatures, surface gravities, metallicity and microturbulent velocity) and equivalent widths of stellar absorption lines. These samples of Ba stars are giants indicated by atmospheric parameters, metallicities and kinematic analysis about UVW velocity. Chemical abundances of 17 elements were obtained for these Ba stars. Their light elements (O, Na, Mg, Al, Si, Ca, Sc, Ti, V, Cr, Mn and Ni) are similar to the solar abundances. Our samples of Ba stars show obvious overabundances of neutron-capture (n-ca...

  1. Applications of Innovative Safety Analysis Methodology (ISAM) to Reload Safety Evaluation

    International Nuclear Information System (INIS)

    KNF has developed the Innovative Safety Analysis Methodology (ISAM) using RETRAN code for Non-LOCA transient analysis during three years from 2006. The first objective of this project is to secure safety analysis methodology required to the export of X-GEN Fuel which KNF is developing. The second is to set up the improved methodology to be applied to the licensing safety analyses for all the OPR1000 and APR1400 plants. The ISAM possesses the characteristics of a designer-friendly methodology. To verify its applicability to the reload safety evaluation, most transients for safety analysis report and for COLSS/CPC setpoint have been analyzed and compared with current safety analysis results. Comparison results show good agreement between them, and it is concluded that the ISAM can be used in the licensing calculations for all the OPR1000 and APR1400 plants. In this paper, presented are the application results of the transients for COLSS/CPC setpoint such as the single CEA withdrawal (SCEAW) event and the asymmetric steam generator transients (ASGT)

  2. Applications of Innovative Safety Analysis Methodology (ISAM) to Reload Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Chan Su; Um, Kil Sup [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2009-05-15

    KNF has developed the Innovative Safety Analysis Methodology (ISAM) using RETRAN code for Non-LOCA transient analysis during three years from 2006. The first objective of this project is to secure safety analysis methodology required to the export of X-GEN Fuel which KNF is developing. The second is to set up the improved methodology to be applied to the licensing safety analyses for all the OPR1000 and APR1400 plants. The ISAM possesses the characteristics of a designer-friendly methodology. To verify its applicability to the reload safety evaluation, most transients for safety analysis report and for COLSS/CPC setpoint have been analyzed and compared with current safety analysis results. Comparison results show good agreement between them, and it is concluded that the ISAM can be used in the licensing calculations for all the OPR1000 and APR1400 plants. In this paper, presented are the application results of the transients for COLSS/CPC setpoint such as the single CEA withdrawal (SCEAW) event and the asymmetric steam generator transients (ASGT)

  3. National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

    International Nuclear Information System (INIS)

    The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

  4. R&D Challenges for SFR Design and Safety Analysis: Opportunities for International Cooperation

    International Nuclear Information System (INIS)

    The paper summarizes the R&D in the safety domain in support of sodium cooled fast reactor (SFR) design and safety analysis. Examples are provided, in particular in the fields of reactivity and decay heat removal control, severe accident analysis, in-service inspection and repair, and chemical risks. It is highlighted that these activities are relevant for international cooperation, especially benchmarks and sharing of experimental facilities. Different frameworks are available for cooperation, such as the Generation IV International Forum, the IAEA (in particular through its coordinated research projects), the European Commission Framework Programme and also bilateral cooperation. (author)

  5. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  6. Safety analysis of the UTSI-CFFF superconducting magnet

    International Nuclear Information System (INIS)

    In designing a large superconducting magnet such as the UTSI-CFFF dipole, great attention must be devoted to the safety of the magnet and personnel. The conductor for the UTSI-CFFF magnet incorporates much copper stabilizer, which both insures its cryostability, and contributes to the magnet safety. The quench analysis and the cryostat fault condition analysis are presented. Two analyses of exposed turns follow; the first shows that gas cooling protects uncovered turns; the second, that the cryostat pressure relief system protects them. Finally the failure mode and safety analysis is presented

  7. Galileo and Ulysses missions safety analysis and launch readiness status

    International Nuclear Information System (INIS)

    The Galileo spacecraft will explore the Jupiter system and Ulysses will fly by Jupiter en route to a polar orbit of the sun. Both spacecraft are powered by general purpose heat source radioisotope thermoelectric generators (RTGs). As a result of the Challenger accident and subsequent mission reprogramming, the Galileo and Ulysses missions' safety analysis had to be repeated. In addition to presenting an overview of the safety analysis status for the missions, this paper presents a brief review of the missions' objectives and design approaches, RTG design characteristics and development history, and a description of the safety analysis process. (author)

  8. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  9. Simulation modeling and analysis in safety. II

    International Nuclear Information System (INIS)

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

  10. An analysis of safety control effectiveness

    International Nuclear Information System (INIS)

    The cost of injuries and 'accidents' to an organisation is very important in establishing how much it should spend on safety control. Despite the usefulness of information about the cost of a company's accidents, it is not customary accounting practice to make these data available. Of the two kinds of costs incurred by a company through occupational injuries and accidents, direct costs and indirect costs; the direct costs are much easier to estimate. However, the uninsured costs are usually more critical and should be estimated by each company. The authors investigate a general model to estimate the above costs and hence to establish efficient safety control. One construction company has been a pilot for this study. By analysing actual company data for three years, it is found that the efficient safety control cost should be 1.2-1.3% of total contract costs

  11. OASIS: An automotive analysis and safety engineering instrument

    International Nuclear Information System (INIS)

    In this paper, we describe a novel software tool named OASIS (AutOmotive Analysis and Safety EngIneering InStrument). OASIS supports automotive safety engineering with features allowing the creation of consistent and complete work products and to simplify and automate workflow steps from early analysis through system development to software development. More precisely, it provides support for (a) model creation and reuse, (b) analysis and documentation and (c) configuration and code generation. We present OASIS as a part of a tool chain supporting the application of a safety engineering workflow aligned with the automotive safety standard ISO 26262. In particular, we focus on OASIS' (1) support for property checking and model correction as well as its (2) support for fault tree generation and FMEA (Failure Modes and Effects Analysis) table generation. Finally, based on the case study of hybrid electric vehicle development, we demonstrate that (1) and (2) are able to strongly support FTA (Fault Tree Analysis) and FMEA

  12. Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

    International Nuclear Information System (INIS)

    This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

  13. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  14. Operational safety analysis status of Novi Han repository

    International Nuclear Information System (INIS)

    This article presents the status of the safety studies and activities related to Novi Han repository. The case of this facility is such that no clear boundary exists between post-closure safety assessment and operational safety assessment. The major findings of these activities are given. The Safety Analysis Report (SAR) for Novi Han repository is developed by Risk Engineering Ltd. under a contract with the Committee on the Use of Atomic Energy for Peaceful Purposes. The general structure and main conclusions and recommendations of the SAR are presented. (author)

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

    International Nuclear Information System (INIS)

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

  16. F-Canyon Suspension and Deactivation Safety Analysis Reports

    International Nuclear Information System (INIS)

    This paper describes Savannah River Site's compliance with the Department of Energy (DOE) direction to suspend current operations, transition to accommodate revised facility missions, and initiate operations to deactivate F-Canyon using a suspension and deactivation safety basis. This paper integrates multiple Workshop theme topics - Lessons Learned from the Safety Analysis Process, Improvements in Documenting Hazard and Accident Analysis, and Closure Issues - Decontamination and Decommissioning. The paper describes the process used to develop safety documentation to support suspension and deactivation activities for F-Canyon. Embodied are descriptive efforts that include development of intermediate and final ''end states'' (e.g., transitional operations), preparation of safety bases documents to support transition, performance of suspension and deactivation activities (e.g. solvent washing, tank/sump flushing, and laboratory waste processing), and downgrade of Safety Class and Safety Significant equipment. The reduction and/or removal of hazards in the facility result in significant risk (frequency times consequence) reduction to the public, site workers, and the environment. Risk reduction then allows the downgrade of safety class and safety significant systems (e.g., ventilation system) and elimination of associated surveillances. The downgrade of safety systems results in significant cost savings

  17. Statistical sampling and chemical analysis of complex weapon components

    International Nuclear Information System (INIS)

    One of the waste streams generated by nuclear weapon dismantlement programs will be component ''hardware'', including complex electronic assemblies such as: radars, arming/fusing/firing systems, power sources, and use-control and safety systems. Sandia National Laboratories (SNL) has been the design and development laboratory for many of these components and will be responsible for their ultimate disposition. This disposition, whether it be reuse, material recycle, or disposal, will require some level of material characterization and analysis. Previous efforts at developing a process for segregation and characterization of hazardous materials in weapon components have been documented. This paper describes the results of recent activities undertaken in support of the Weapon Hardware Inventory Reduction Effort (WHIRE) at Sandia National Laboratories. These activities have been directed principally towards: The development of a statistically sound sampling plan for chemical analysis of weapon component materials; the development of a non-destructive analytical screening method for determining the Toxicity Characteristic of excess weapon hardware

  18. Safety analysis and evaluation methodology for fusion systems

    International Nuclear Information System (INIS)

    A synthesized methodology of safety analysis and evaluation for general fusion systems is proposed. In the course of the methodology development, its main frame has been constructed in order to take account of all safety-related items and to ensure a logical consistency. The safety-related items are divided broadly into two groups. One of them is the public protection from radiological hazard, which is introduced as a safety requirement from an external viewpoint for the fusion system. The other items are the matter from an internal viewpoint and are related to the fusion system behavior in itself. These items are composed of the understanding of a fusion system, the safety ensuring principle and the function based safety analysis. All of these items have been mapped on the frame, considering the mutual relations, among them, consistently. To complete the methodology development, the safety evaluation for the actual design of a fusion system has been performed in conformity to this methodology. Thus, it has been demonstrated that the methodology proposed here is appropriate to the safety analysis and evaluation for the fusion system. (author). 9 refs, 4 figs, 2 tabs

  19. Testing Chemical Safety: What Is Needed to Ensure the Widespread Application of Non-animal Approaches?

    Directory of Open Access Journals (Sweden)

    Natalie Burden

    2015-05-01

    Full Text Available Scientists face growing pressure to move away from using traditional animal toxicity tests to determine whether manufactured chemicals are safe. Numerous ethical, scientific, business, and legislative incentives will help to drive this shift. However, a number of hurdles must be overcome in the coming years before non-animal methods are adopted into widespread practice, particularly from regulatory, scientific, and global perspectives. Several initiatives are nevertheless underway that promise to increase the confidence in newer alternative methods, which will support the move towards a future in which less data from animal tests is required in the assessment of chemical safety.

  20. Testing Chemical Safety: What Is Needed to Ensure the Widespread Application of Non-animal Approaches?

    Science.gov (United States)

    Burden, Natalie; Sewell, Fiona; Chapman, Kathryn

    2015-05-01

    Scientists face growing pressure to move away from using traditional animal toxicity tests to determine whether manufactured chemicals are safe. Numerous ethical, scientific, business, and legislative incentives will help to drive this shift. However, a number of hurdles must be overcome in the coming years before non-animal methods are adopted into widespread practice, particularly from regulatory, scientific, and global perspectives. Several initiatives are nevertheless underway that promise to increase the confidence in newer alternative methods, which will support the move towards a future in which less data from animal tests is required in the assessment of chemical safety. PMID:26018957

  1. Automation of Safety Analysis with SysML Models Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This project was a small proof-of-concept case study, generating SysML model information as a side effect of safety analysis. A prototype FMEA Assistant was...

  2. Safety analysis methodologies for radioactive waste repositories in shallow ground

    International Nuclear Information System (INIS)

    The report is part of the IAEA Safety Series and is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of shallow ground radioactive waste repositories. It discusses approaches that are applicable for safety analysis of a shallow ground repository. The methodologies, analysis techniques and models described are pertinent to the task of predicting the long-term performance of a shallow ground disposal system. They may be used during the processes of selection, confirmation and licensing of new sites and disposal systems or to evaluate the long-term consequences in the post-sealing phase of existing operating or inactive sites. The analysis may point out need for remedial action, or provide information to be used in deciding on the duration of surveillance. Safety analysis both general in nature and specific to a certain repository, site or design concept, are discussed, with emphasis on deterministic and probabilistic studies

  3. Development and improvement of safety analysis code for geological disposal

    International Nuclear Information System (INIS)

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  4. West Valley Reprocessing Plant. Safety analysis report, supplement 21

    International Nuclear Information System (INIS)

    Supplement No. 21 contains responses to USNRC questions on quality assurance contained in USNRC letter to NFS dated January 22, 1976, revised pages for the safety analysis report, and Appendix IX ''Quality Assurance Manual--West Valley Construction Projects.''

  5. Analysis of Fouling Resistance in Safety-Related Heat Exchangers

    International Nuclear Information System (INIS)

    As a part of nuclear safety activities, developed countries have performed Periodic Safety Review (PSR) to verify and improve the safety of operating Nuclear Power Plants (NPPs). In 1999, it was decided by the Korean Atomic Energy Safety Committee to adopt the PSR program. PSR is officially legislated in 2001 as a 10-year-basis safety evaluation process. Since the first tentative application of PSR for Gori Unit 1 in 2000, it is now progressing well. Generally PSR assesses the cumulative effects of plant ageing and plant modifications, operating experience, technical developments and site aspects. The reviews include an assessment of plant design and operation against current safety standards and practices. After reviewing activities, safety is enhanced by implementing the corrective actions and/or safety improvements. When a PSR was performed in Gori Units 3-4, several safety-related heat exchangers in the Reactor Coolant System (RCS) such as a letdown heat exchanger were pointed out as the components necessitating a corrective action which is the analysis of fouling resistance. The fouling resistance is used as an important parameter to evaluate the safety as well as the economics of heat exchangers. However it is difficult to develop a credible analysis procedure due to considerable discrepancy between normal operating conditions and design conditions. This issue was identified while we were conducting a study in KNICS (Korea Nuclear I and C System) R and D program. We might be able to guess other NPPs in Korea are likely to have the same issue. This paper involves the characteristics of the safety related heat exchangers and the methodology to develop the analysis procedure

  6. Quantitative Safety and Security Analysis from a Communication Perspective

    Directory of Open Access Journals (Sweden)

    Boris Malinowsky

    2015-12-01

    Full Text Available This paper introduces and exemplifies a trade-off analysis of safety and security properties in distributed systems. The aim is to support analysis for real-time communication and authentication building blocks in a wireless communication scenario. By embedding an authentication scheme into a real-time communication protocol for safety-critical scenarios, we can rely on the protocol’s individual safety and security properties. The resulting communication protocol satisfies selected safety and security properties for deployment in safety-critical use-case scenarios with security requirements. We look at handover situations in a IEEE 802.11 wireless setup between mobile nodes and access points. The trade-offs involve application-layer data goodput, probability of completed handovers, and effect on usable protocol slots, to quantify the impact of security from a lower-layer communication perspective on the communication protocols. The results are obtained using the network simulator ns-3.

  7. Quantitative Safety and Security Analysis from a Communication Perspective

    DEFF Research Database (Denmark)

    Malinowsky, Boris; Schwefel, Hans-Peter; Jung, Oliver

    2014-01-01

    This paper introduces and exemplifies a trade-off analysis of safety and security properties in distributed systems. The aim is to support analysis for real-time communication and authentication building blocks in a wireless communication scenario. By embedding an authentication scheme into a real......-time communication protocol for safety-critical scenarios, we can rely on the protocol’s individual safety and security properties. The resulting communication protocol satisfies selected safety and security properties for deployment in safety-critical use-case scenarios with security requirements. We look at...... handover situations in a IEEE 802.11 wireless setup between mobile nodes and access points. The trade-offs involve application-layer data goodput, probability of completed handovers, and effect on usable protocol slots, to quantify the impact of security from a lower-layer communication perspective on the...

  8. Statistical margin to DNB safety analysis approach for LOFT

    International Nuclear Information System (INIS)

    A method was developed and used for LOFT thermal safety analysis to estimate the statisticl margin to DNB for the hot rod, and to base safety analysis on desired DNB probability limits. This method is an advanced approach using response surface analysis methods, a very efficient experimental design, and a 2nd-order response surface equation with a 2nd-order error propagation analysis to define the MDNBR probability density function. Calculations for limiting transients were used in the response surface analysis thereby including transient interactions and trip uncertainties in the MDNBR probability density

  9. SAFETY ANALYSIS OF THE DEMONSTRATION BULK VITRIFICATION SYSTEM

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) and CH2M HILL, Hanford Group, Inc. (CH2M HILL) [also referred to as the Tank Farm Contractor (TFC)] are evaluating the Demonstration Bulk Vitrification System (DBVS) as a supplemental treatment technology for low-activity waste (LAW) at the Hanford Site. As a new facility at Hanford, the safety analysis for the DBVS is being subjected to new and evolving DOE requirements. Hazard categorization for the facility is being closely examined since this determines whether performance category (PC)-2 or PC-3 requirements are to be applied for natural phenomena hazards, as well as differing requirements under Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2004-2, Active Confinement Systems. Questions have also arisen regarding application of DOE-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, or DOE-STD-1189-2008, Integration of Safety into the Design Process, format and content, as well as full implementation of DOE-STD-1186-2004, Specific Administrative Controls and naming conventions and content requirements for the interim safety analysis documents under DOE O 413.3A, Program and Project Management for the Acquisition of Capital Assets (e.g., Preliminary Documented Safety Analysis or Preliminary Safety Design Report). Another challenge is the integration of the programmatic chapters of the safety document with those of the Hanford tank farms, since the tank farms Safety Management Programs (SMP) are relied upon for the DBVS facility. All of these issues and their resolutions, as well as the level of scrutiny to which internal and external regulators have held this project's safety analysis, will be discussed in this paper

  10. Using Qualitative Hazard Analysis to Guide Quantitative Safety Analysis

    Science.gov (United States)

    Shortle, J. F.; Allocco, M.

    2005-01-01

    Quantitative methods can be beneficial in many types of safety investigations. However, there are many difficulties in using quantitative m ethods. Far example, there may be little relevant data available. This paper proposes a framework for using quantitative hazard analysis to prioritize hazard scenarios most suitable for quantitative mziysis. The framework first categorizes hazard scenarios by severity and likelihood. We then propose another metric "modeling difficulty" that desc ribes the complexity in modeling a given hazard scenario quantitatively. The combined metrics of severity, likelihood, and modeling difficu lty help to prioritize hazard scenarios for which quantitative analys is should be applied. We have applied this methodology to proposed concepts of operations for reduced wake separation for airplane operatio ns at closely spaced parallel runways.

  11. Analysis of Chemical Technology Division waste streams

    International Nuclear Information System (INIS)

    This document is a summary of the sources, quantities, and characteristics of the wastes generated by the Chemical Technology Division (CTD) of the Oak Ridge National Laboratory. The major contributors of hazardous, mixed, and radioactive wastes in the CTD as of the writing of this document were the Chemical Development Section, the Isotopes Section, and the Process Development Section. The objectives of this report are to identify the sources and the summarize the quantities and characteristics of hazardous, mixed, gaseous, and solid and liquid radioactive wastes that are generated by the Chemical Technology Division (CTD) of the Oak Ridge National Laboratory (ORNL). This study was performed in support of the CTD waste-reduction program -- the goals of which are to reduce both the volume and hazard level of the waste generated by the division. Prior to the initiation of any specific waste-reduction projects, an understanding of the overall waste-generation system of CTD must be developed. Therefore, the general approach taken in this study is that of an overall CTD waste-systems analysis, which is a detailed presentation of the generation points and general characteristics of each waste stream in CTD. The goal of this analysis is to identify the primary waste generators in the division and determine the most beneficial areas to initiate waste-reduction projects. 4 refs., 4 figs., 13 tabs

  12. Recent Progresses in Nanobiosensing for Food Safety Analysis.

    Science.gov (United States)

    Yang, Tao; Huang, Huifen; Zhu, Fang; Lin, Qinlu; Zhang, Lin; Liu, Junwen

    2016-01-01

    With increasing adulteration, food safety analysis has become an important research field. Nanomaterials-based biosensing holds great potential in designing highly sensitive and selective detection strategies necessary for food safety analysis. This review summarizes various function types of nanomaterials, the methods of functionalization of nanomaterials, and recent (2014-present) progress in the design and development of nanobiosensing for the detection of food contaminants including pathogens, toxins, pesticides, antibiotics, metal contaminants, and other analytes, which are sub-classified according to various recognition methods of each analyte. The existing shortcomings and future perspectives of the rapidly growing field of nanobiosensing addressing food safety issues are also discussed briefly. PMID:27447636

  13. SCALE 5: Powerful new criticality safety analysis tools

    International Nuclear Information System (INIS)

    Version 5 of the SCALE computer software system developed at Oak Ridge National Laboratory, scheduled for release in December 2003, contains several significant new modules and sequences for criticality safety analysis and marks the most important update to SCALE in more than a decade. This paper highlights the capabilities of these new modules and sequences, including continuous energy flux spectra for processing multigroup problem-dependent cross sections; one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations; two-dimensional flexible mesh discrete ordinates code; automated burnup-credit analysis sequence; and one-dimensional material distribution optimization for criticality safety. (author)

  14. Recent Progresses in Nanobiosensing for Food Safety Analysis

    Science.gov (United States)

    Yang, Tao; Huang, Huifen; Zhu, Fang; Lin, Qinlu; Zhang, Lin; Liu, Junwen

    2016-01-01

    With increasing adulteration, food safety analysis has become an important research field. Nanomaterials-based biosensing holds great potential in designing highly sensitive and selective detection strategies necessary for food safety analysis. This review summarizes various function types of nanomaterials, the methods of functionalization of nanomaterials, and recent (2014–present) progress in the design and development of nanobiosensing for the detection of food contaminants including pathogens, toxins, pesticides, antibiotics, metal contaminants, and other analytes, which are sub-classified according to various recognition methods of each analyte. The existing shortcomings and future perspectives of the rapidly growing field of nanobiosensing addressing food safety issues are also discussed briefly. PMID:27447636

  15. Construction safety and waste management an economic analysis

    CERN Document Server

    Li, Rita Yi Man

    2015-01-01

    This monograph presents an analysis of construction safety problems and on-site safety measures from an economist’s point of view. The book includes examples from both emerging countries, e.g. China and India, and developed countries, e.g. Australia and Hong Kong. Moreover, the author covers an analysis on construction safety knowledge sharing by means of updatable mobile technology such as apps in Androids and iOS platform mobile devices. The target audience comprises primarily researchers and experts in the field but the book may also be beneficial for graduate students.

  16. Computational methods for criticality safety analysis within the scale system

    International Nuclear Information System (INIS)

    The criticality safety analysis capabilities within the SCALE system are centered around the Monte Carlo codes KENO IV and KENO V.a, which are both included in SCALE as functional modules. The XSDRNPM-S module is also an important tool within SCALE for obtaining multiplication factors for one-dimensional system models. This paper reviews the features and modeling capabilities of these codes along with their implementation within the Criticality Safety Analysis Sequences (CSAS) of SCALE. The CSAS modules provide automated cross-section processing and user-friendly input that allow criticality safety analyses to be done in an efficient and accurate manner. 14 refs., 2 figs., 3 tabs

  17. Unavailability analysis of redundant safety systems

    International Nuclear Information System (INIS)

    Analytical equations have been obtained for the unavailabilities of redundant standby safety systems with components tested periodically. Test and repair contributions, hardware failures, human testing and repair errors as well as failures due to true demands have been taken into account. Equations have been derived for m-out-of-n systems (1 less than or equal to m less than or equal to n less than or equal to 4) with uniformly staggered, consecutive and random testing schemes. The equations have been used in a computer code, ICARUS, and applied to practical safety systems. The results are useful for optimizing the redundancy and testing and they illustrate the importance of human/testing errors and falures associated with true demands

  18. Components, Safety Interfaces, and Compositional Analysis

    OpenAIRE

    Elmqvist, Jonas

    2007-01-01

    Component-based software development has emerged as a promising approach for developing complex software systems by composing smaller independently developed components into larger component assemblies. This approach offers means to increase software reuse, achieve higher flexibility and shorter time-to-market by the use of off-the-shelf components (COTS). However, the use of COTS in safety-critical system is highly unexplored. This thesis addresses the problems appearing in component-based d...

  19. Entrainment analysis and monitoring major safety systems

    International Nuclear Information System (INIS)

    The authors are convinced that taking account of internal and external experience and a plant-specific living PSA frequently reduces the notifiable incidents occurring as design errors due to inadequate checks on safety margins. On the basis of the considerations formulated in this article, Leibstadt nuclear power station has decided to overhaul the earlier PSA and work towards and implement a living PSA. The project has been given the green light and should be completed in two years. 5 figs., 4 refs

  20. 14 CFR 417.309 - Flight safety system analysis.

    Science.gov (United States)

    2010-01-01

    ... 12-dB margin, each link analysis must account for the following nominal system performance and... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Flight safety system analysis. 417.309... analysis. (a) General. (1) Each flight termination system and command control system, including each...

  1. Chemical conditions in present and future ecosystems in Forsmark - implications for selected radionuclides in the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Troejbom, Mats (Mats Troejbom Konsult AB (Sweden)); Grolander, Sara (Facilia AB (Sweden))

    2010-12-15

    This report is a background report for the biosphere analysis of the SR-Site Safety Assessment. This work aims to describe the future development of the chemical conditions at Forsmark, based on the present chemical conditions at landscape level taking landscape development and climate cases into consideration. The results presented contribute to the overall understanding of the present and future chemistry in the Forsmark area, and specifically, to the understanding of the behaviour of some selected radionuclides in the surface system. The future development of the chemistry at the site is qualitatively discussed with focus on the interglacial within the next 10,000 years. The effects on the chemical environment of future climate cases as Global Warming and cold permafrost climates are also briefly discussed. The work is presented in two independent parts describing background radionuclide activities in the Forsmark area and the distribution and behaviour of a large number of stable elements in the landscape. In a concluding section, implications of the future chemical environment of a selection of radionuclides important in the Safety Assessment are discussed based on the knowledge of stable elements. The broad range of elements studied show that there are general and expected patterns for the distribution and behaviour in the landscape of different groups of elements. Mass balances reveal major sources and sinks, pool estimations show where elements are accumulated in the landscape and estimations of time-scales give indications of the potential future development. This general knowledge is transferred to radionuclides not measured in order to estimate their behaviour and distribution in the landscape. It could be concluded that the future development of the chemical environment in the Forsmark area might affect element specific parameters used in de radionuclide model in different directions depending on element. The alternative climate cases, Global Warming

  2. Probabilistic safety analysis for the Unterweser Nuclear Power Station

    International Nuclear Information System (INIS)

    In October last year, a plant-specific probabilistic safety analysis (PSA) was conducted for the Unterweser Nuclear Power Station as part of the periodic safety review (PSR). As a living PSA, the probabilistic safety analysis was based on the first analysis conducted in 1995; its scope was extended in accordance with the 1996 PSA guideline. Besides the in-plant initiating events in the power mode, which were considered already in the 1995 PSA, the current PSA included external impacts, fires in the plant, and events occurring during plant outages as well as plant-specific data. Also findings of current research were incorporated. The results obtained show the KKU plant to enjoy a high level of safety and allow the PSA to be used alongside plant operation. (orig.)

  3. Exploring the limits of safety analysis in complex technological systems

    CERN Document Server

    Sornette, D; Kroeger, W

    2012-01-01

    From biotechnology to cyber-risks, most extreme technological risks cannot be reliably estimated from historical statistics. Engineers resort to probability safety analysis (PSA), which consists in developing models to simulate accidents, potential scenarios, their severity and frequency. However, even the best safety analysis struggles to account for evolving risks resulting from inter-connected networks and cascade effects. Taking nuclear risks as an example, the predicted plant-specific distribution of losses is found to be significantly underestimated when compared with available empirical records. A simple cascade model suggests that the classification of the different possible safety regimes is intrinsically unstable in the presence of cascades. Even the best probabilistic safety analysis requires additional continuous validation, making the best use of the experienced realized incidents, near misses and accidents.

  4. Westinghouse Hanford Company safety analysis reports and technical safety requirements upgrade program

    International Nuclear Information System (INIS)

    During Fiscal Year 1992, the US Department of Energy, Richland Operations Office (RL) separately transmitted the following US Department of Energy (DOE) Orders to Westinghouse Hanford Company (WHC) for compliance: DOE 5480.21, ''Unreviewed Safety Questions,'' DOE 5480.22, ''Technical Safety Requirements,'' and DOE 5480.23, ''Nuclear Safety Analysis Reports.'' WHC has proceeded with its impact assessment and implementation process for the Orders. The Orders are closely-related and contain some requirements that are either identical, similar, or logically-related. Consequently, WHC has developed a strategy calling for an integrated implementation of the three Orders. The strategy is comprised of three primary objectives, namely: Obtain DOE approval of a single list of DOE-owned and WHC-managed Nuclear Facilities, Establish and/or upgrade the ''Safety Basis'' for each Nuclear Facility, and Establish a functional Unreviewed Safety Question (USQ) process to govern the management and preservation of the Safety Basis for each Nuclear Facility. WHC has developed policy-revision and facility-specific implementation plans to accomplish near-term tasks associated with the above strategic objectives. This plan, which as originally submitted in August 1993 and approved, provided an interpretation of the new DOE Nuclear Facility definition and an initial list of WHC-managed Nuclear Facilities. For each current existing Nuclear Facility, existing Safety Basis documents are identified and the plan/status is provided for the ISB. Plans for upgrading SARs and developing TSRs will be provided after issuance of the corresponding Rules

  5. Aquatic environmental safety assessment and inhibition mechanism of chemicals for targeting Microcystis aeruginosa.

    Science.gov (United States)

    Yu, Xiao-Bo; Hao, Kai; Ling, Fei; Wang, Gao-Xue

    2014-11-01

    Cyanobacteria are a diverse group of Gram-negative bacteria that produce an array of secondary compounds with selective bioactivity against vertebrates, invertebrates, fungi, bacteria and cell lines. Recently the main methods of controlling cyanobacteria are using chemicals, medicinal plants and microorganism but fewer involved the safety research in hydrophytic ecosystems. In search of an environmentally safe compound, 53 chemicals were screened against the developed heavy cyanobacteria bloom Microcystis aeruginosa using coexistence culture system assay. The results of the coexistence assay showed that 9 chemicals inhibited M. aeruginosa effectively at 20 mg L(-1) after 7 days of exposure. Among them dimethomorph, propineb, and paraquat were identified that they are safe for Chlorella vulgaris, Scenedesmus obliquus, Carassius auratus (Goldfish) and Bacillus subtilis within half maximal effective concentration (EC50) values 5.2, 4.2 and 0.06 mg L(-1) after 7 days, respectively. Paraquat as the positive control observed to be more efficient than the other compounds with the inhibitory rate (IR) of 92% at 0.5 mg L(-1). For the potential inhibition mechanism, the chemicals could destroy the cell ultrastructure in different speed. The safety assay proved dimethomorph, propineb and paraquat as harmless formulations or products having potential value in M. aeruginosa controlling, with the advantage of its cell morphology degrading ability. PMID:25139029

  6. VALIDATION GUIDELINES FOR LABORATORIES PERFORMING FORENSIC ANALYSIS OF CHEMICAL TERRORISM

    Science.gov (United States)

    The Scientific Working Group on Forensic Analysis of Chemical Terrorism (SWGFACT) has developed the following guidelines for laboratories engaged in the forensic analysis of chemical evidence associated with terrorism. This document provides a baseline framework and guidance for...

  7. CHEMICALS

    CERN Multimedia

    Medical Service

    2002-01-01

    It is reminded that all persons who use chemicals must inform CERN's Chemistry Service (TIS-GS-GC) and the CERN Medical Service (TIS-ME). Information concerning their toxicity or other hazards as well as the necessary individual and collective protection measures will be provided by these two services. Users must be in possession of a material safety data sheet (MSDS) for each chemical used. These can be obtained by one of several means : the manufacturer of the chemical (legally obliged to supply an MSDS for each chemical delivered) ; CERN's Chemistry Service of the General Safety Group of TIS ; for chemicals and gases available in the CERN Stores the MSDS has been made available via EDH either in pdf format or else via a link to the supplier's web site. Training courses in chemical safety are available for registration via HR-TD. CERN Medical Service : TIS-ME :73186 or service.medical@cern.ch Chemistry Service : TIS-GS-GC : 78546

  8. PSA analysis focused on Mochovce NPP safety measures evaluation from operational safety point of view

    International Nuclear Information System (INIS)

    Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present two units are operated and another two are under the construction. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on the exploitation of similar units. These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for SMs establishing including their grouping covering different areas of safety goals and results of SM contributions to the total core damage frequency based on FPSA analysis. (author)

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

    Science.gov (United States)

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

    2015-12-01

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

  10. Probabilistic safety analysis and radiological protection

    International Nuclear Information System (INIS)

    The author presents a brief description of NUREG-1150 and NUREG-0956, both documents of great importance in the risk area. Based on document's recommendations and following NUREG-1150 similar methodology, a calculation model is proposed in this publication, with the purpose of analyzing the consequences of a severe accident in Angra-I Power Station. The suggested model can be divided in two stages: the first one called front-end considers the power station system safety during the accident, and the second called back-end cares for accident consequences. 9 refs. (B.C.A.)

  11. Safety Analysis of Stochastic Dynamical Systems

    DEFF Research Database (Denmark)

    Sloth, Christoffer; Wisniewski, Rafael

    2015-01-01

    This paper presents a method for verifying the safety of a stochastic system. In particular, we show how to compute the largest set of initial conditions such that a given stochastic system is safe with probability p. To compute the set of initial conditions we rely on the moment method that via...... Haviland's theorem allows an infinite dimensional optimization problem on measures to be formulated as a polynomial optimization problem. Subsequently, the moment sequence is truncated (relaxed) to obtain a finite dimensional polynomial optimization problem. Finally, we provide an illustrative example that...

  12. Safety Analysis of Liquid Rocket Engine Using Bayesian Networks

    Institute of Scientific and Technical Information of China (English)

    WANG Hua-wei; YAN Zhi-qiang

    2007-01-01

    Safety analysis for liquid rocket engine has a great meaning for shortening development cycle, saving development expenditure and reducing development risk. The relationship between the structure and component of liquid rocket engine is much more complex, furthermore test data are absent in development phase. Thereby, the uncertainties exist in safety analysis for liquid rocket engine. A safety analysis model integrated with FMEA(failure mode and effect analysis)based on Bayesian networks (BN) is brought forward for liquid rocket engine, which can combine qualitative analysis with quantitative decision. The method has the advantages of fusing multi-information, saving sample amount and having high veracity. An example shows that the method is efficient.

  13. Evaluation of atmospheric dispersion/consequence models supporting safety analysis

    International Nuclear Information System (INIS)

    Two DOE Working Groups have completed evaluation of accident phenomenology and consequence methodologies used to support DOE facility safety documentation. The independent evaluations each concluded that no one computer model adequately addresses all accident and atmospheric release conditions. MACCS2, MATHEW/ADPIC, TRAC RA/HA, and COSYMA are adequate for most radiological dispersion and consequence needs. ALOHA, DEGADIS, HGSYSTEM, TSCREEN, and SLAB are recommended for chemical dispersion and consequence applications. Additional work is suggested, principally in evaluation of new models, targeting certain models for continued development, training, and establishing a Web page for guidance to safety analysts

  14. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

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

  15. A Safety Analysis Approach to Clinical Workflows: Application and Evaluation

    Directory of Open Access Journals (Sweden)

    Lamis Al-Qora’n

    2014-11-01

    Full Text Available Clinical workflows are safety critical workflows as they have the potential to cause harm or death to patients. Their safety needs to be considered as early as possible in the development process. Effective safety analysis methods are required to ensure the safety of these high-risk workflows, because errors that may happen through routine workflow could propagate within the workflow to result in harmful failures of the system’s output. This paper shows how to apply an approach for safety analysis of clinical workflows to analyse the safety of the workflow within a radiology department and evaluates the approach in terms of usability and benefits. The outcomes of using this approach include identification of the root causes of hazardous workflow failures that may put patients’ lives at risk. We show that the approach is applicable to this area of healthcare and is able to present added value through the detailed information on possible failures, of both their causes and effects; therefore, it has the potential to improve the safety of radiology and other clinical workflows.

  16. Safety analysis and evaluation methodology for fusion systems

    International Nuclear Information System (INIS)

    Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

  17. Express method of nuclear safety analysis for VVER

    International Nuclear Information System (INIS)

    The original criterion method of nuclear safety analysis for WWER with Western nuclear fuel is presented. The method is based on adequate interdependence of safety criteria for fuel matrix and fuel cladding, and on conservative phenomenological criteria for nuclear fuel heat release power and heat transfer conditions in specific accident scenarios. The tolerability criteria for the temperature of nuclear fuel and fuel cladding from zirconium is analysed. The method based on the conservative criteria for analysis of nuclear safety is proposed. The heat balance equation and the boundary conditions of the external heat exchange are derived. The criteria for the safety for the temperature of fuel rod and cladding was obtained. The proposed method don’t require modeling of all possible accident sequences using detailed codes. Therefore, scope of computational studies are essentially reduced. In addition, it enables fast adaptation of criterion method for express-evaluation of the nuclear safety variations for different initial events and conditions, and at modification and/or change of nuclear fuel design. Keywords: nuclear safety, safety criterion, nuclear fuel, water-moderated water-cooled reactor (WWER), heat exchange

  18. Safety analysis of JAEA-ADS in Japan

    International Nuclear Information System (INIS)

    It is considered that the ADS is safer than conventional critical nuclear reactors because the operation of the ADS is able to be stopped by the shutdown of the accelerator. Meanwhile, it is important to comprehend transients at the failure of the proton beam shutdown for the discussion of the ADS inherent safety. To investigate transients at such unprotected accidents, safety analyses for the JAEA ADS were performed. SIMMER-III, an advanced safety analysis code, was employed for the safety analysis. As typical accident scenarios, following two cases were analyzed by SIMMER-III code; unprotected beam overpower (UBOP) and unprotected loss of flow (ULOF). In this study, the word ‘unprotected’ means no shutdown of the proton beam during severe sequences 

  19. Preliminary safety analysis report for the Waste Characterization Facility

    International Nuclear Information System (INIS)

    This safety analysis report outlines the safety concerns associated with the Waste Characterization Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are to: define and document a safety basis for the Waste Characterization Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume. 142 refs., 38 figs., 39 tabs

  20. Safety Analysis and Risk Assessment Handbook, new guidance to the safety analyst

    International Nuclear Information System (INIS)

    New guidance to the safety analyst has been developed at the Rocky Flats Environmental Technology Site (RFETS) in the form of the new Safety Analysis and Risk Assessment Handbook (SARAH). Although the older guidance (the Rocky Flats Risk Assessment Guide) continues to be used for updating the Final Safety Analysis Reports (FSARs) developed in the mid-1980s, this new guidance is used with all new authorization basis documents. With the RFETS mission change in the early 1990s came the need to establish new authorization basis documents for its facilities, whose missions had changed. The methodology and databases for performing the evaluations that support the new authorization basis documents needed to be standardized, to avoid the use of different approaches and/or databases for similar accidents in different facilities. This paper presents this new standardized approach, the SARAH

  1. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  2. Exploring the limits of safety analysis in complex technological systems

    OpenAIRE

    Sornette, D.; Maillart, T.; Kroeger, W.

    2012-01-01

    From biotechnology to cyber-risks, most extreme technological risks cannot be reliably estimated from historical statistics. Therefore, engineers resort to predictive methods, such as fault/event trees in the framework of probabilistic safety assessment (PSA), which consists in developing models to identify triggering events, potential accident scenarios, and estimate their severity and frequency. However, even the best safety analysis struggles to account for evolving risks resulting from in...

  3. Safety analysis report for packaging (onsite) steel drum

    Energy Technology Data Exchange (ETDEWEB)

    McCormick, W.A.

    1998-09-29

    This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum.

  4. Safety analysis report for packaging (onsite) steel drum

    International Nuclear Information System (INIS)

    This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum

  5. 242-A evaporator safety analysis report

    International Nuclear Information System (INIS)

    In compliance with DOE Orders, an update of the 242-A SAR has been prepared, as documented in the referenced ECN. Several categories of changes were identified for inclusion in this revision of the SAR. These categories will be utilized to simplify the discussion of the changes for this USQ document. However, it is important to note that no new tests or experiments were included in this revision of the SAR. Editorial changes and/or informational updates to Chapters 9 and 11 were included as part of this revision. However, no changes to Operational Safety Requirements (OSRs) contained in Chapter 11 were required. General categories of changes included in this revision are listed

  6. Cosmetics chemical composition characterization by instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Brazil is in the third position in the world's cosmetics market. It is an expanding and growing market where new products and manufacturing processes are in a constant and steady expansion. Therefore, it is mandatory that the composition of the products is well known in order to guarantee safety and quality of daily used cosmetics. The Brazilian National Health Surveillance Agency (ANVISA) has issued a resolution, RDC No. 48, March 16, 2006, which defines a 'List of Substances which can not be used in personal hygiene products, cosmetics and perfumes'. In this work, samples of locally manufactured and imported cosmetics (lipsticks, eye shadows, etc.) were analyzed using the Instrumental Neutron Activation Analysis technique. The samples were irradiated in the TRIGA IPR-R1 reactor of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN), on a 100kW thermal power, with a thermal neutron fluence rate about 8x1011ncm-2s-1. The analysis has detected the chemical elements Br, Ba, Ga, Na, K, Sc, Fe, Cr, Zn, Sm, W, La, Rb, Cs, Ta, Ge, Co, U, Ti, V, Cl, Al, Mn and Cu. The concentrations of these elements are on a range from 5 to 3000μg.g-1. Some chemical elements observed in samples (Cl, Br, Cr, U) are included at ANVISA prohibitive list. (author)

  7. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  8. Reactivity insertion assumptions used in safety analysis calculations

    International Nuclear Information System (INIS)

    The report discusses the basis for selection of the trip reactivity insertion curve used by B and W in the safety analysis of its nuclear steam supply systems. The implementation of new three-dimensional calculational techniques and test results from operating plants now allow derivation of reactivity insertion curves that vary about the previously symmetric insertion curve used in the safety analysis. The report is generically applicable to all operating or nearly operating 177-FA (fuel assembly) plants in its evaluation of the impact of these new trip reactivity insertion curves. In conjunction with the TMI-2 FSAR review, the NRC expressed concern over the scram times defined by the Technical Specifications and how they were related to the safety analysis presented in Chapter 15. This concern is discussed in the report, and a proposed Technical Specification change is presented that is responsive to the staff's concern. The control rod drive mechanism (CRDM) is discussed, and the results of on-site tests are presented. Calculated integral rod worth data are compared to measured data. Also discussed are factors that influence the shape of this curve, such as flux redistribution and initial off-set. The trip reactivity insertion curve used in the safety analysis is discussed. This curve is a composite of the rod velocity curve and the integral rod worth curve. The synthesis of two curves to produce the reactivity insertion curve is illustrated. The impact of the resultant trip reactivity insertion curve on the safety analysis is also described

  9. LWR core safety analysis with Areva's 3-dimensional methods

    International Nuclear Information System (INIS)

    The quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools and an extensive validation base. Sophisticated 3-dimensional core models ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. The validation base includes measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models achieve reliable and comprehensive results for a wide range of applications. As an example an overview of the application experience as well as the validation base of AREVA's 3-dimensional codes is given. The importance and necessity of the comprehensive 3-dimensional methodology is illustrated with examples of a BWR and PWR safety analysis. For BWR transient application the analysis of regional power oscillations is considered and regarding the PWR safety analysis an example referring to fast enthalpy rise and the maximum fuel temperature caused by a rod ejection accident is shown. (orig.)

  10. Waste Sampling & Characterization Facility (WSCF) Complex Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    MELOY, R.T.

    2002-04-01

    This document was prepared to analyze the Waste Sampling and Characterization Facility for safety consequences by: Determining radionuclide and highly hazardous chemical inventories; Comparing these inventories to the appropriate regulatory limits; Documenting the compliance status with respect to these limits; and Identifying the administrative controls necessary to maintain this status. The primary purpose of the Waste Sampling and Characterization Facility (WSCF) is to perform low-level radiological and chemical analyses on various types of samples taken from the Hanford Site. These analyses will support the fulfillment of federal, Washington State, and Department of Energy requirements.

  11. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

  12. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

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

  13. The Efficacy of a Condensed "Seeking Safety" Intervention for Women in Residential Chemical Dependence Treatment at 30 Days Posttreatment

    Science.gov (United States)

    Cash Ghee, Anna; Bolling, Lanny C.; Johnson, Candace S.

    2009-01-01

    This study examined the efficacy of a condensed version of the "Seeking Safety" intervention in the reduction of trauma-related symptoms and improved drug abstinence rates among women in residential chemical dependence treatment. One hundred and four women were randomly assigned to treatment including a condensed (six session) "Seeking Safety"…

  14. Systems analysis of past, present, and future chemical terrorism scenarios.

    Energy Technology Data Exchange (ETDEWEB)

    Hoette, Trisha Marie

    2012-03-01

    Throughout history, as new chemical threats arose, strategies for the defense against chemical attacks have also evolved. As a part of an Early Career Laboratory Directed Research and Development project, a systems analysis of past, present, and future chemical terrorism scenarios was performed to understand how the chemical threats and attack strategies change over time. For the analysis, the difficulty in executing chemical attack was evaluated within a framework of three major scenario elements. First, historical examples of chemical terrorism were examined to determine how the use of chemical threats, versus other weapons, contributed to the successful execution of the attack. Using the same framework, the future of chemical terrorism was assessed with respect to the impact of globalization and new technologies. Finally, the efficacy of the current defenses against contemporary chemical terrorism was considered briefly. The results of this analysis justify the need for continued diligence in chemical defense.

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

    International Nuclear Information System (INIS)

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

  16. Principal Component Analysis on Chemical Abundances Spaces

    CERN Document Server

    Ting, Y S; Kobayashi, C; De Silva, G M; Bland-Hawthorn, J

    2011-01-01

    [Shortened] In preparation for the HERMES chemical tagging survey of about a million Galactic FGK stars, we estimate the number of independent dimensions of the space defined by the stellar chemical element abundances [X/Fe]. [...] We explore abundances in several environments, including solar neighbourhood thin/thick disk stars, halo metal-poor stars, globular clusters, open clusters, the Large Magellanic Cloud and the Fornax dwarf spheroidal galaxy. [...] We find that, especially at low metallicity, the production of r-process elements is likely to be associated with the production of alpha-elements. This may support the core-collapse supernovae as the r-process site. We also verify the over-abundances of light s-process elements at low metallicity, and find that the relative contribution decreases at higher metallicity, which suggests that this lighter elements primary process may be associated with massive stars. [...] Our analysis reveals two types of core-collapse supernovae: one produces mainly alpha-e...

  17. Methodological basis for analysis and accounting of NPP probabilistic safety analysis uncertainties

    International Nuclear Information System (INIS)

    The paper presents classification of NPP probabilistic safety analysis uncertainties and defines their main sources. It sets forth methods to perform statistical and analytical analysis of different uncertainty classes, proposes sequence of efforts related to analysis and accounting of uncertainties in making decisions on NPP safety

  18. Chemical safety of cassava products in regions adopting cassava production and processing - experience from Southern Africa

    DEFF Research Database (Denmark)

    Nyirenda, D.B.; Chiwona-Karltun, L.; Chitundu, M.;

    2011-01-01

    perceptions concerning cassava and chemical food safety. Chips, mixed biscuits and flour, procured from households and markets in three regions of Zambia (Luapula-North, Western and Southern) as well as products from the Northern, Central and Southern regions of Malawi, were analyzed for total cyanogenic...... products commercially available on the market. Risk assessments disclose that effects harmful to the developing central nervous system (CNS) may be observed at a lower exposure than previously anticipated. We interviewed farmers in Zambia and Malawi about their cultivars, processing procedures and...

  19. Safety analysis of the existing 851 Firing Facility

    International Nuclear Information System (INIS)

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 851 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but two of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exceptions were the linear accelerator and explosives, which were classified as moderate hazards per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  20. Safety analysis of the existing 850 Firing Facility

    International Nuclear Information System (INIS)

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 850 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives, which was classified as a moderate hazard per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  1. Possibilities of Moessbauer spectroscopy for chemical analysis

    International Nuclear Information System (INIS)

    Full text: The Moessbauer spectroscopy technique belongs to few methods of defining the phase state or crystallographic sites of a substance. The Moessbauer spectra bear information on various hyperfine interactions, many of which are indirectly related to the chemical nature of the Moessbauer atom and its nearest environment. Determination of the parameters of hyperfine interactions that can be extracted from Moessbauer spectra and used for qualitative analysis is a routine task. In the present work, we studied the influence of various factors on experimental errors encountered in quantitatively defining the phase composition or site populations of the substance under study, such as the measurements geometry, Lamb-Moessbauer coefficients, absorber thickness, efficiency and dead time of the detection system and spectral line shape. The absolute f measurements were made using the 'black' absorber method. Moessbauer measurements were carried out with carefully controlled background intensities, since the accuracy of f evaluation directly depends on the measurement of the background. The influence of a non-uniformity of samples on the results of the quantitative analysis is discussed. The data analysis was divided into two parts: removal of instrumental artifacts by folding and baseline correction and deconvolution to extract the hyperfine parameters of individual local environments. In our approach, calibration graphs were drawn by measuring the spectra of a series of analogous samples having different known concentrations. For the same purpose, the internal standard method was also used. Experimental data are presented for phase analyses of different samples. (author)

  2. Advanced development in chemical analysis of Cordyceps.

    Science.gov (United States)

    Zhao, J; Xie, J; Wang, L Y; Li, S P

    2014-01-01

    Cordyceps sinensis, also called DongChongXiaCao (winter worm summer grass) in Chinese, is a well-known and valued traditional Chinese medicine. In 2006, we wrote a review for discussing the markers and analytical methods in quality control of Cordyceps (J. Pharm. Biomed. Anal. 41 (2006) 1571-1584). Since then this review has been cited by others for more than 60 times, which suggested that scientists have great interest in this special herbal material. Actually, the number of publications related to Cordyceps after 2006 is about 2-fold of that in two decades before 2006 according to the data from Web of Science. Therefore, it is necessary to review and discuss the advanced development in chemical analysis of Cordyceps since then. PMID:23688494

  3. LOCA analysis of SCWR-M with passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

    2013-06-15

    Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

  4. AP1000 passive safety system design and analysis

    International Nuclear Information System (INIS)

    Westinghouse Electric Company has designed an advanced 600 MW nuclear power plant called the AP600. The AP600 uses passive safety systems to enhance plant safety and to satisfy US licensing requirements. The use of passive safety systems has provided significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. The overnight capital cost for the first AP600 plant is calculated to be between 1300- 1500 $/kw depending on the site selection. Although the AP600 is the most cost effective plant ready for deployment, it is still more expensive than the $1000/kw needed to compete in the United States today. In order to develop a cost competitive nuclear power plant Westinghouse has completed design studies which demonstrate that it is feasible to increase the power output of the AP600 to at least 1000 MW, maintaining its current design configuration, use of proven components and licensing basis. The AP1000 reactor and passive safety features retain the same configuration as the AP600. The approach to designing the passive core cooling features is to evaluate each feature to determine if changes are necessary to provide proper safety margins at the higher power rating. Both design basis and PRA based accidents sequences are considered in this evaluation. Insights from the extensive AP600 test and analysis program are used to assist in this process. The results of preliminary accident analysis for DBA and PRA sequences are used to demonstrate the effectiveness of this approach. (author)

  5. Analysis for design of passive safety injection line in IPSS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jihee; Kim, Sangho; Chang, Soonheung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    The current safety system of nuclear power plants cannot deal with loss-of-coolant accidents during the circumstance of station black-out (SBO), total loss of AC electric power. However, application of IPSS allows nuclear power plants to solve the combined accidents by its characteristics, only operated by natural phenomena. In order to achieve ultimate safety from the IPSS, analysis for currently operating nuclear power plants should be considered. Hence, in this research, analysis for the effectiveness of passive safety injection of IPSS was conducted for OPR1000 with using MARS code. In this study, application of PSIS for OPR1000 was evaluated. Following the results were simulated by MARS, we verified that PSIS can be the successful supplement for safety roles that HPSI and LPSI does in OPR1000. Even though, the result is strictly bounded to OPR1000, it gives us a prospect that PSIS of IPSS can cope with LBLOCA in the failure of active safety systems induced by SBO on other PWR. On the other hand, this simulation was evaluated with a hypothesis of direct vessel injection. Therefore, further study is needed for passive safety injection through cold legs to OPR1000 and other PWRs and those comparisons.

  6. Safety analysis for small and medium size integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jin H.; Chang, M. H.; Bae, K. H.; Lim, H. S.; Kwon, M.; Lee, Y. J.; Hwang, Y. D.; Kim, S. O.; Lee, W. J.; Chung, B. D.

    1997-09-01

    Sets of safety and performance related design basis events have been proposed for the SMART. Detailed descriptions of the events, justification and the selection criteria are specified. Operation modes of the SMART integral reactor are described. Safety systems as well as the components specific to the SMART integral reactor are evaluated. Thermal hydraulic system codes are evaluated for the use of the safety and performance analysis. Both the safety and performance methodology as well as the code systems are proposed for the safety and performance analysis of the SMART integral reactor. A preliminary PIRT for the SMART integral reactor was developed by an expert panel during the study. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect test was developed for the thermal hydraulic model development and the system code validation. This experimental program will be also used to evaluated the safety systems and to support licensing confirmation of the SMART integral reactor. The results of the study will be used for the conceptual design of the SMART integral reactor. (author). 58 refs., 23 tabs., 30 figs.

  7. Capillary electrophoresis for the analysis of contaminants in emerging food safety issues and food traceability.

    Science.gov (United States)

    Vallejo-Cordoba, Belinda; González-Córdova, Aarón F

    2010-07-01

    This review presents an overview of the applicability of CE in the analysis of chemical and biological contaminants involved in emerging food safety issues. Additionally, CE-based genetic analyzers' usefulness as a unique tool in food traceability verification systems was presented. First, analytical approaches for the determination of melamine and specific food allergens in different foods were discussed. Second, natural toxin analysis by CE was updated from the last review reported in 2008. Finally, the analysis of prion proteins associated with the "mad cow" crises and the application of CE-based genetic analyzers for meat traceability were summarized. PMID:20593390

  8. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  9. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  10. Advanced analysis and design for fire safety of steel structures

    CERN Document Server

    Li, Guoqiang

    2013-01-01

    Advanced Analysis and Design for Fire Safety of Steel Structures systematically presents the latest findings on behaviours of steel structural components in a fire, such as the catenary actions of restrained steel beams, the design methods for restrained steel columns, and the membrane actions of concrete floor slabs with steel decks. Using a systematic description of structural fire safety engineering principles, the authors illustrate the important difference between behaviours of an isolated structural element and the restrained component in a complete structure under fire conditions. The book will be an essential resource for structural engineers who wish to improve their understanding of steel buildings exposed to fires. It is also an ideal textbook for introductory courses in fire safety for master’s degree programs in structural engineering, and is excellent reading material for final-year undergraduate students in civil engineering and fire safety engineering. Furthermore, it successfully bridges th...

  11. 2014 PGSFR Safety Analysis for Loss of Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. H.; Lee, K. L.; Choi, C. W.; Jeong, T. K.; Yoo, J.; Chang, W. P.; Ahn, S. J.; Lee, S. W.; Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The PGSFR consists of the PHTS (Primary Heat Transport System), the IHTS (Intermediate Heat Transport System), and the DHRS (Decay Heat Removal System). A LOF (Loss Of Flow) accident has been investigated for a safety evaluation of the PGSFR using the MARSLMR code. The safety analysis is evaluated by a CDF (Cumulative Damage Fraction). In case of the LOF accident, the tentative safety criterion is the CDF of under 0.05. The LOF accident has been evaluated in the PGSFR using MARS-LMR. The accident was initiated by both of PHTS pump trip. In the results, the CDF was predicted below a tentative safety criterion of 0.05 with a sufficient margin. The DHRS acceptably functioned for removing the core decay heat during long-term cooling period.

  12. Safety analysis of SISL process module

    International Nuclear Information System (INIS)

    This report provides an assessment of various postulated accidental occurrences within an experimental process module which is part of a Special Isotope Separation Laboratory (SISL) currently under construction at the Lawrence Livermore National Laboratory (LLNL). The process module will contain large amounts of molten uranium and various water-cooled structures within a vacuum vessel. Special emphasis is therefore given to potential accidental interactions of molten uranium with water leading to explosive and/or rapid steam formation, as well as uranium oxidation and the potential for combustion. Considerations are also given to the potential for vessel melt-through. Evaluations include mechanical and thermal interactions and design implications both in terms of design basis as well as once-in-a-lifetime accident scenarios. These scenarios include both single- and multiple-failure modes leading to various contact modes and locations within the process module for possible thermal interactions. The evaluations show that a vacuum vessel design based upon nominal operating conditions would appear sufficient to meet safety requirements in connection with both design basis as well as once-in-a-lifetime accidents. Controlled venting requirements for removal of steam and hydrogen in order to avoid possible long-term pressurization events are recommended. Depending upon the resulting accident conditions, the vacuum system (i.e., the roughing system) could also serve this purpose. Finally, based upon accident evaluations of this study, immediate shut-off of all coolant water following an incident leak is not recommended, as such action may have adverse effects in terms of cool-down requirements for the melt crucibles etc. These requirements have not been assessed as part of this study

  13. Comprehensive safety analysis for pressure and cryogenic systems facilities

    International Nuclear Information System (INIS)

    There have been many instances where serious injuries and fatalities have resulted from over-pressurization, thermal stress, asphyxiation and other potential hazards associated with testing, handling and storage of compressed gases and cryogenic liquids at numerous production and research facilities. These hazards are major issues that should be addressed in system design and in materials selection appropriate for high pressure or cryogenic temperature applications. Potential hazards may be mitigated through system analysis and design process which are the major factors in preventing thermal/pressure hazards caused by possible leaks and fragmentation, in the case of rupture. This paper presents a conceptual model and framework for developing a comprehensive safety analysis which will reduce potential hazards, accidents and legal liabilities. The proposed in-depth system Safety Analysis Report (SAR) is a proven systematic approach to identify hazards and influence design to provide timely documentation of potential hazards and risks associated with systems, facilities, and equipment. As a result of this hazard analysis process, provisions and actions for hazard prevention, elimination, mitigation, and control have been put in place, and all identifiable potential hazards have been reduced to a low risk level. These methods are demonstrated in the example of comprehensive safety analysis of Cryogenic Subsystem of Accelerator String Test facilities (ASST) at Superconducting Super Collider Laboratory by developing Safety Analysis Report (SSC Laboratory, 1992)

  14. Use of safety analysis results to support process operation

    International Nuclear Information System (INIS)

    Safety and risk analysis carried out during the design phase of a process plant produces useful knowledge about the behavior and the disturbances of the system. This knowledge, however, often remains to the designer though it would be of benefit to the operators and supervisors of the process plant, too. In Technical Research Centre of Finland a project has been started to plan and construct a prototype of an information system to make use of the analysis knowledge during the operation phase. The project belongs to a Nordic KRM project (Knowledge Based Risk Management System). The information system is planned to base on safety and risk analysis carried out during the design phase and completed with operational experience. The safety analysis includes knowledge about potential disturbances, their causes and consequences in the form of Hazard and Operability Study, faut trees and/or event trees. During the operation disturbances can however, occur, which are not included in the safety analysis, or the causes or consequences of which have been incompletely identified. Thus the information system must also have an interface for the documentation of the operational knowledge missing from the analysis results. The main tasks off the system when supporting the management of a disturbance are to identify it (or the most important of the coexistent ones) from the stored knowledge and to present it in a proper form (for example as a deviation graph). The information system may also be used to transfer knowledge from one shift to another and to train process personnel

  15. Contribution of simplified vehicle dynamic models to road safety analysis

    OpenAIRE

    ORFILA, O; VANDANJON, PO; COIRET, A

    2008-01-01

    The skid resistance analysis is one part of the complex process which is involved in road safety analysis for a road project or for road maintenance. As early as the design phase, the road geometry choice is linked to the skid resistance for a given level of service. In the road maintenance case, different conventional measurements are performed on the road and by analyzing them together, risks can be enlightened. Currently, this analysis is done by comparing these data. This is time consumin...

  16. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  17. Safety analysis for boiling water reactors

    International Nuclear Information System (INIS)

    This report is the translation of GRS-95 'Sicherheitsanalyse fuer Siedewasserreaktoren - Zusammenfassende Darstellung'. Recent analysis results -concerning the chapters on accident management, fire and earthquake - that were not included in the German text have been added to this translation. In cases of doubt, GRS-102 (main volume) is the factually correct version. (orig.)

  18. Development of evaluation method for software safety analysis techniques

    International Nuclear Information System (INIS)

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

  19. Safety analysis and review system: a Department of Energy safety assurance tool

    International Nuclear Information System (INIS)

    The concept of the Safety Analysis and Review System is not new. It has been used within the Department and its predecessor agencies, Atomic Energy Commission (AEC) and Energy Research and Development Administration (ERDA), for over 20 years. To minimize the risks from nuclear reactor and power plants, the AEC developed a process to support management authorization of each operation through identification and analysis of potential hazards and the measures taken to control them. As the agency evolved from AEC through ERDA to the Department of Energy, its responsibilities were broadened to cover a diversity of technologies, including those associated with the development of fossil, solar, and geothermal energy. Because the safety analysis process had proved effective in a technology of high potential hazard, the Department investigated the applicability of the process to the other technologies. This paper describes the system and discusses how it is implemented within the Department

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

    DEFF Research Database (Denmark)

    The report describes presentations and discussions at a seminar held at Risø on March 22-23, 2000. The title of the seminar was NKS/SOS-1 – Safety Analysis. It dealt with issues of relevance for the safety analysis for the entire nuclear safety field (notably reactors and nuclear waste repositories......). Such issues were: objectives of safety analysis, risk criteria, decision analysis, expert judgement and risk communication. In addition, one talk dealt with criteria for chemical industries in Europe. The seminar clearly showed that the concept of risk is multi-dimensional, which makes clarity and...

  1. Safety and evidence concept. Report on the work package 4. Preliminary safety analysis for the site Gorleben

    International Nuclear Information System (INIS)

    The national legal regulations including the German atomic law, the radiation protection law and the federal mining law define the general framework conditions for the final disposal of radioactive waste. In addition the safety requirements concerning heat generating radioactive wastes are of importance. The project VSG (preliminary safety analysis for the site Gorleben) is aimed to the safety of the installation following the closure phase of the repository. The safety concept discusses the legal requirement, the evidence concept is concerned with the demonstration of the efficacy of the safety concept with respect to the safe enclosure of radionuclides and the assessment of long-term safety based on model calculations.

  2. Tolerance and safety of superficial chemical peeling with salicylic acid in various facial dermatoses

    Directory of Open Access Journals (Sweden)

    Iqbal Zafar

    2005-03-01

    Full Text Available BACKGROUND: Chemical peeling is a skin-wounding procedure that may have some potentially undesirable side-effects. AIMS: The present study is directed towards safety concerns associated with superficial chemical peeling with salicylic acid in various facial dermatoses. METHODS: The study was a non-comparative and a prospective one. Two hundred and sixty-eight patients of either sex, aged between 10 to 60 years, undergoing superficial chemical peeling for various facial dermatoses (melasma, acne vulgaris, freckles, post-inflammatory scars/pigmentation, actinic keratoses, plane facial warts, etc. were included in the study. Eight weekly peeling sessions were carried out in each patient. Tolerance to the procedure and any undesirable effects noted during these sessions were recorded. RESULTS: Almost all the patients tolerated the procedure well. Mild discomfort, burning, irritation and erythema were quite common but the incidence of major side-effects was very low and these too, were easily manageable. There was no significant difference in the incidence of side-effects between facial dermatoses (melasma, acne and other pigmentary disorders. CONCLUSION: Chemical peeling with salicylic acid is a well tolerated and safe treatment modality in many superficial facial dermatoses.

  3. Tools to prevent process safety events at university research facility - chemical risk assessment and experimental set-up risk assessment

    DEFF Research Database (Denmark)

    Jensen, Niels; Jørgensen, Sten Bay

    2014-01-01

    The article discusses the two forms developed to examine the hazards of the chemicals to be used in the experiments in the experimental setup in the Department of Chemical and Biochemical Engineering of the Technical University of Denmark. A system for the safety assessment of new experimental...

  4. Nuclear criticality safety evaluation -- DWPF Late Wash Facility, Salt Process Cell and Chemical Process Cell

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) High Level Nuclear Waste will be vitrified in the Defense Waste Processing Facility (DWPF) for long term storage and disposal. This is a nuclear criticality safety evaluation for the Late Wash Facility (LWF), the Salt Processing Cell (SPC) and the Chemical Processing Cell (CPC). of the DWPF. Waste salt solution is processed in the Tank Farm In-Tank Precipitation (ITP) process and is then further washed in the DWPF Late Wash Facility (LWF) before it is fed to the DWPF Salt Processing Cell. In the Salt Processing Cell the precipitate slurry is processed in the Precipitate Reactor (PR) and the resultant Precipitate Hydrolysis Aqueous (PHA) produce is combined with the sludge feed and frit in the DWPF Chemical Process Cell to produce a melter feed. The waste is finally immobilized in the Melt Cell. Material in the Tank Farm and the ITP and Extended Sludge processes have been shown to be safe against a nuclear criticality by others. The precipitate slurry feed from ITP and the first six batches of sludge feed are safe against a nuclear criticality and this evaluation demonstrates that the processes in the LWF, the SPC and the CPC do not alter the characteristics of the materials to compromise safety

  5. Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

    International Nuclear Information System (INIS)

    This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 711 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition2. Results of the analysis and testing performed on the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of energy (DOE) Order 5480.33 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.94 and 7.10.5

  6. Safety analysis report - packages 9965, 9968, 9972-9975 packages

    International Nuclear Information System (INIS)

    This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B( ), 9968 B( ), 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 10 CFR 71 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition. Results of the analysis and testing performed on the 9965 B(), 9968 B(), 9972 B(U), 9973 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of Energy (DOE) Order 5480.3 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.9 and 7.10

  7. Final Safety Analysis Report (FSAR) for Building 332, Increment III

    Energy Technology Data Exchange (ETDEWEB)

    Odell, B. N.; Toy, Jr., A. J.

    1977-08-31

    This Final Safety Analysis Report (FSAR) supplements the Preliminary Safety Analysis Report (PSAR), dated January 18, 1974, for Building 332, Increment III of the Plutonium Materials Engineering Facility located at the Lawrence Livermore Laboratory (LLL). The FSAR, in conjunction with the PSAR, shows that the completed increment provides facilities for safely conducting the operations as described. These documents satisfy the requirements of ERDA Manual Appendix 6101, Annex C, dated April 8, 1971. The format and content of this FSAR complies with the basic requirements of the letter of request from ERDA San to LLL, dated March 10, 1972. Included as appendices in support of th FSAR are the Building 332 Operational Safety Procedure and the LLL Disaster Control Plan.

  8. Risk and safety analysis of nuclear systems

    CERN Document Server

    Lee, John C

    2011-01-01

    The book has been developed in conjunction with NERS 462, a course offered every year to seniors and graduate students in the University of Michigan NERS program. The first half of the book covers the principles of risk analysis, the techniques used to develop and update a reliability data base, the reliability of multi-component systems, Markov methods used to analyze the unavailability of systems with repairs, fault trees and event trees used in probabilistic risk assessments (PRAs), and failure modes of systems. All of this material is general enough that it could be used in non-nuclear a

  9. Human Pluripotent Stem Cell Based Developmental Toxicity Assays for Chemical Safety Screening and Systems Biology Data Generation.

    Science.gov (United States)

    Shinde, Vaibhav; Klima, Stefanie; Sureshkumar, Perumal Srinivasan; Meganathan, Kesavan; Jagtap, Smita; Rempel, Eugen; Rahnenführer, Jörg; Hengstler, Jan Georg; Waldmann, Tanja; Hescheler, Jürgen; Leist, Marcel; Sachinidis, Agapios

    2015-01-01

    Efficient protocols to differentiate human pluripotent stem cells to various tissues in combination with -omics technologies opened up new horizons for in vitro toxicity testing of potential drugs. To provide a solid scientific basis for such assays, it will be important to gain quantitative information on the time course of development and on the underlying regulatory mechanisms by systems biology approaches. Two assays have therefore been tuned here for these requirements. In the UKK test system, human embryonic stem cells (hESC) (or other pluripotent cells) are left to spontaneously differentiate for 14 days in embryoid bodies, to allow generation of cells of all three germ layers. This system recapitulates key steps of early human embryonic development, and it can predict human-specific early embryonic toxicity/teratogenicity, if cells are exposed to chemicals during differentiation. The UKN1 test system is based on hESC differentiating to a population of neuroectodermal progenitor (NEP) cells for 6 days. This system recapitulates early neural development and predicts early developmental neurotoxicity and epigenetic changes triggered by chemicals. Both systems, in combination with transcriptome microarray studies, are suitable for identifying toxicity biomarkers. Moreover, they may be used in combination to generate input data for systems biology analysis. These test systems have advantages over the traditional toxicological studies requiring large amounts of animals. The test systems may contribute to a reduction of the costs for drug development and chemical safety evaluation. Their combination sheds light especially on compounds that may influence neurodevelopment specifically. PMID:26132533

  10. Chemical abundance analysis of 19 barium stars

    Science.gov (United States)

    Yang, Guo-Chao; Liang, Yan-Chun; Spite, Monique; Chen, Yu-Qin; Zhao, Gang; Zhang, Bo; Liu, Guo-Qing; Liu, Yu-Juan; Liu, Nian; Deng, Li-Cai; Spite, Francois; Hill, Vanessa; Zhang, Cai-Xia

    2016-01-01

    We aim at deriving accurate atmospheric parameters and chemical abundances of 19 barium (Ba) stars, including both strong and mild Ba stars, based on the high signal-to-noise ratio and high resolution Echelle spectra obtained from the 2.16 m telescope at Xinglong station of National Astronomical Observatories, Chinese Academy of Sciences. The chemical abundances of the sample stars were obtained from an LTE, plane-parallel and line-blanketed atmospheric model by inputting the atmospheric parameters (effective temperatures Teff, surface gravities log g, metallicity [Fe/H] and microturbulence velocity ξt) and equivalent widths of stellar absorption lines. These samples of Ba stars are giants as indicated by atmospheric parameters, metallicities and kinematic analysis about UVW velocity. Chemical abundances of 17 elements were obtained for these Ba stars. Their Na, Al, α- and iron-peak elements (O, Na, Mg, Al, Si, Ca, Sc, Ti, V, Cr, Mn, Ni) are similar to the solar abundances. Our samples of Ba stars show obvious overabundances of neutron-capture (n-capture) process elements relative to the Sun. Their median abundances of [Ba/Fe], [La/Fe] and [Eu/Fe] are 0.54, 0.65 and 0.40, respectively. The Y I and Zr I abundances are lower than Ba, La and Eu, but higher than the α- and iron-peak elements for the strong Ba stars and similar to the iron-peak elements for the mild stars. There exists a positive correlation between Ba intensity and [Ba/Fe]. For the n-capture elements (Y, Zr, Ba, La), there is an anti-correlation between their [X/Fe] and [Fe/H]. We identify nine of our sample stars as strong Ba stars with [Ba/Fe] >0.6 where seven of them have Ba intensity Ba=2-5, one has Ba=1.5 and another one has Ba=1.0. The remaining ten stars are classified as mild Ba stars with 0.17<[Ba/Fe] <0.54.

  11. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

  12. An Empirical Analysis of Human Performance and Nuclear Safety Culture

    International Nuclear Information System (INIS)

    The purpose of this analysis, which was conducted for the US Nuclear Regulatory Commission (NRC), was to test whether an empirical connection exists between human performance and nuclear power plant safety culture. This was accomplished through analyzing the relationship between a measure of human performance and a plant's Safety Conscious Work Environment (SCWE). SCWE is an important component of safety culture the NRC has developed, but it is not synonymous with it. SCWE is an environment in which employees are encouraged to raise safety concerns both to their own management and to the NRC without fear of harassment, intimidation, retaliation, or discrimination. Because the relationship between human performance and allegations is intuitively reciprocal and both relationship directions need exploration, two series of analyses were performed. First, human performance data could be indicative of safety culture, so regression analyses were performed using human performance data to predict SCWE. It also is likely that safety culture contributes to human performance issues at a plant, so a second set of regressions were performed using allegations to predict HFIS results

  13. Spectroscopic Chemical Analysis Methods and Apparatus

    Science.gov (United States)

    Hug, William F.; Reid, Ray D.

    2012-01-01

    This invention relates to non-contact spectroscopic methods and apparatus for performing chemical analysis and the ideal wavelengths and sources needed for this analysis. It employs deep ultraviolet (200- to 300-nm spectral range) electron-beam-pumped wide bandgap semiconductor lasers, incoherent wide bandgap semiconductor lightemitting devices, and hollow cathode metal ion lasers. Three achieved goals for this innovation are to reduce the size (under 20 L), reduce the weight [under 100 lb (.45 kg)], and reduce the power consumption (under 100 W). This method can be used in microscope or macroscope to provide measurement of Raman and/or native fluorescence emission spectra either by point-by-point measurement, or by global imaging of emissions within specific ultraviolet spectral bands. In other embodiments, the method can be used in analytical instruments such as capillary electrophoresis, capillary electro-chromatography, high-performance liquid chromatography, flow cytometry, and related instruments for detection and identification of unknown analytes using a combination of native fluorescence and/or Raman spectroscopic methods. This design provides an electron-beampumped semiconductor radiation-producing method, or source, that can emit at a wavelength (or wavelengths) below 300 nm, e.g. in the deep ultraviolet between about 200 and 300 nm, and more preferably less than 260 nm. In some variations, the method is to produce incoherent radiation, while in other implementations it produces laser radiation. In some variations, this object is achieved by using an AlGaN emission medium, while in other implementations a diamond emission medium may be used. This instrument irradiates a sample with deep UV radiation, and then uses an improved filter for separating wavelengths to be detected. This provides a multi-stage analysis of the sample. To avoid the difficulties related to producing deep UV semiconductor sources, a pumping approach has been developed that uses

  14. International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

    International Nuclear Information System (INIS)

    The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

  15. Behaviour analysis of AC-600 passive safety systems

    International Nuclear Information System (INIS)

    Southwest Center of Reactor Engineering Research and Design has finished the first step conceptual design of 600 mwe advanced PWR (AC-600). The main research emphases of AC-600 conceptual design include the advanced reactor core, the passive safety systems and the simplification. The passive safety systems of AC-600 consist of two reactor make up water tanks, two accumulators, two emergency feedwater tanks, two emergency natural draft air condensers, a containment water jacket and an enhanced primary cycle natural circulation flow system. 25% of the rated reactor power can be removed by the natural circulation cooling. The full pressure reactor make up water tanks are able to provide enough borated water which would be injected into the reactor coolant system during small LOCA. The coolant natural circulations can be established in the primary system and the passive secondary emergency feedwater system, removing residual heat from the reactor core to the atmosphere when station blackout occurs. It is indicated from analysis that the containment diameter of AC-600 is about 35 m. The large tanks and the large vertical distances between the tanks and reactor core are the main reason of using the big containment. It is also indicated from analysis that the low head safety injection pumps are required in AC-600 design to assure the recirculation system operation when large LOCA occurs. The reliability of AC-600 engineered safety systems is increased because the function of the passive safety systems is conducted through the immutable natural laws. The paper discusses the natural circulation ability and safety behavior of the passive safety systems during LOCA or station blackout for AC-600. The passive limits to excess reactivity and thermal hydraulic transients are also preliminarily discussed. Figs and tabs

  16. CONACS: the DOE safety analysis system

    International Nuclear Information System (INIS)

    The CONtainment Analysis Code System (CONACS) is a large, comprehensive scientific simulation system for predicting conditions in an LMR facility following the occurrence of a postulated accident. It has now been developed to a stage of completion that can be referred to as a limited operational version. This version forms a permanent portion of the ultimate system. Because CONACS was developed with change in mind, it is now possible to draw on this strength to respond to changing requirements arising from advanced design concepts. The generalized design applications in the nuclear and non-nuclear fields and the quality assurance applied to the project make those adaptations reliable. In this paper the results of prototype tests and the implications of limited version tests are presented along with a brief description of CONACS and its relationship to LMR design optimization and cost reduction

  17. CONACS: the DOE safety analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Martin, F.J.; Armstrong, G.R.; Niccoli, L.G.

    1985-03-01

    The CONtainment Analysis Code System (CONACS) is a large, comprehensive scientific simulation system for predicting conditions in an LMR facility following the occurrence of a postulated accident. It has now been developed to a stage of completion that can be referred to as a limited operational version. This version forms a permanent portion of the ultimate system. Because CONACS was developed with change in mind, it is now possible to draw on this strength to respond to changing requirements arising from advanced design concepts. The generalized design applications in the nuclear and non-nuclear fields and the quality assurance applied to the project make those adaptations reliable. In this paper the results of prototype tests and the implications of limited version tests are presented along with a brief description of CONACS and its relationship to LMR design optimization and cost reduction.

  18. Analysis of safety margins for PuO2 containers

    International Nuclear Information System (INIS)

    In the regular manner the containers for PuO2 transport are type B(U) and give satisfaction to the AIEA proofs. However the vigour of this conception's containers and the analysis of other radioactive containers permit to think that large safety margins exist. In this paper, the importance and the kind of these margins are studied

  19. 10 CFR 72.248 - Safety analysis report updating.

    Science.gov (United States)

    2010-01-01

    ... Approval of Spent Fuel Storage Casks § 72.248 Safety analysis report updating. (a) Each certificate holder for a spent fuel storage cask design shall update periodically, as provided in paragraph (b) of this... Commission, in accordance with § 72.4, within 90 days after the spent fuel storage cask design has...

  20. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

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

  1. Safety analysis and the code development on radioactive waste disposal

    International Nuclear Information System (INIS)

    Regarding development of the safety analysis codes to be used for 'cross-check' (which is the evaluation of the validity of the safety analysis conducted by the licensee through cross comparison of the simulated result) of the sub-surface disposal conducted by the licensee, the codes are required to be capable of confirming the long term safety of the sub-surface disposal. The influence of the rainfall infiltration change on groundwater flow over the long term period due to climate change was studied. As a result, it was found that shoreline movement caused by the sea level change significantly influenced groundwater flow. Regarding development of the safety analysis codes to be used for 'cross-check' of the near surface disposal, it is important to efficiently simulate the groundwater flow with finely discretized mesh model. We therefore improved the memory allocation algorithm of the groundwater flow simulation code, TOUGH2 to be able to treat the large mesh model, such as several million cells. Modifications are made for the simulation support system, by adding the groundwater flow code 3D-SEEP which can treat land uplift and erosion and its associated modules. This modification not only improves efficiency but also allows to avoid human error. Moreover, sensitivity analysis of the unsaturated conditions such as infiltration rate on the migration of important nuclides of near surface disposal was conducted. As a result, influence of the unsaturated conditions on the exposed dose was evaluated. (author)

  2. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    A standard model for a safety analysis report of fuel fabrication plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.)

  3. Chernobyl ''Sarcophagus'' safety analysis and proposals for modernization

    International Nuclear Information System (INIS)

    Safety analysis and proposals for modernization of Chernobyl ''Sarcophagus'' are presented, including the following aspects: the state of Unit 4 after the accident; construction of the ''Sarcophagus''; inspections; structure state monitoring; transformation of the facility into a long-term safe system

  4. Fast flux test facility final safety analysis report amendment 79

    International Nuclear Information System (INIS)

    This document is provided to replace, remove, or add applicable pages to the chapters on: Heat Transport System; Containment and Structures; Auxiliary Systems; Reactor Refueling System; Conduct of Operations; Safety Analysis; Quality Assurance; FFTF Criticality Specifications; and Appendix H's TRIGA Fuel Storage System

  5. Standard model for safety analysis report of fuel reprocessing plants

    International Nuclear Information System (INIS)

    A standard model for a safety analysis report of fuel reprocessing plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.)

  6. QuantUM: Quantitative Safety Analysis of UML Models

    Directory of Open Access Journals (Sweden)

    Florian Leitner-Fischer

    2011-07-01

    Full Text Available When developing a safety-critical system it is essential to obtain an assessment of different design alternatives. In particular, an early safety assessment of the architectural design of a system is desirable. In spite of the plethora of available formal quantitative analysis methods it is still difficult for software and system architects to integrate these techniques into their every day work. This is mainly due to the lack of methods that can be directly applied to architecture level models, for instance given as UML diagrams. Also, it is necessary that the description methods used do not require a profound knowledge of formal methods. Our approach bridges this gap and improves the integration of quantitative safety analysis methods into the development process. All inputs of the analysis are specified at the level of a UML model. This model is then automatically translated into the analysis model, and the results of the analysis are consequently represented on the level of the UML model. Thus the analysis model and the formal methods used during the analysis are hidden from the user. We illustrate the usefulness of our approach using an industrial strength case study.

  7. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  8. Safety criteria and guidelines for MSR accident analysis

    International Nuclear Information System (INIS)

    Accident analysis for Molten Salt Reactor (MSR) has been investigated at ORNL for MSRE in 1960s. Since then, safety criteria or guidelines have not been defined for MSR accident analysis. Regarding the safety criteria, the authors showed one proposal in this paper. In order to establish guidelines for MSR accident analysis, we have to investigate all possible accidents. In this paper, the authors describe the philosophy for accident analysis, and show 40 possible accidents. They are at first classified as external cause accidents and internal cause accidents. Since the former ones are generic accidents, we investigate only the latter ones, and categorize them to 4 types, such as power excursion accident, flow decrease accident, fuel-salt leak accident, and other accidents mostly specific to MSR. Each accident is described briefly, with some numerical results by the authors. (author)

  9. Synthesis report for the VSG (preliminary safety analysis Gorleben). Report on working package 13. Preliminary safety analysis for the Gorleben site

    International Nuclear Information System (INIS)

    The report on the preliminary safety analysis for the Gorleben site covers the following issues: Accomplishment of the project covering the 14 working packages: fundamentals; geosciences on the description of the site; waste specification and quantity structure, safety and verification concept, development of final repository concepts, repository design and optimization, evaluation of human intrusion, considerations on the operational security, system analysis, FEP (features, events and processes) catalogue, development of scenarios, integrity analysis, radiological consequence analysis, synthesis and recommendations. Special emphasis is given to the topics safety and verification concept, realization of the safety and verification concept in the frame of the preliminary safety analysis for the Gorleben site.

  10. PAT-1 safety analysis report addendum.

    Energy Technology Data Exchange (ETDEWEB)

    Weiner, Ruth F.; Schmale, David T.; Kalan, Robert J.; Akin, Lili A.; Miller, David Russell; Knorovsky, Gerald Albert; Yoshimura, Richard Hiroyuki; Lopez, Carlos; Harding, David Cameron; Jones, Perry L.; Morrow, Charles W.

    2010-09-01

    The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The purpose of this SAR Addendum is to incorporate plutonium (Pu) metal as a new payload for the PAT-1 package. The Pu metal is packed in an inner container (designated the T-Ampoule) that replaces the PC-1 inner container. The documentation and results from analysis contained in this addendum demonstrate that the replacement of the PC-1 and associated packaging material with the T-Ampoule and associated packaging with the addition of the plutonium metal content are not significant with respect to the design, operating characteristics, or safe performance of the containment system and prevention of criticality when the package is subjected to the tests specified in 10 CFR 71.71, 71.73 and 71.74.

  11. Structural safety analysis of HTGR core supports

    Energy Technology Data Exchange (ETDEWEB)

    Ju, F.; Bennett, J.G.; Anderson, C.A.

    1977-01-01

    In the current design of the High Temperature Gas-Cooled Reactor (HTGR), the core is made up of stacked columns of graphite fuel blocks. Structural support for the core takes the form of graphite columns beneath the core together with lateral springs, which position and restrain the core from contact with the sides of the reactor containment vessel. Each individual support column carries the dead load of several fuel columns together with the equivalent load caused by the coolant pressure drop through the core. The adequacy of the support structure to provide torsional stability of the core for both static and seismic loadings as well as long term stability of the core support structure itself is discussed. Analysis for long term stability of the core support columns involves consideration of eccentric loading (caused by damaged spherical seats) and imperfections in the form of surface cracks. Nonlinear graphite behavior must also be taken into consideration. For predictions of the core torsional seismic response, the core was represented as a right circular cylinder supported on elastic posts; the lateral support was represented by a single torsional spring. Energy losses from friction and material hysteresis were represented by viscous dampers. The coupled equations for vertical and rotational motions were integrated numerically and dynamic core response was computed fromtorsional acceleration time-histories obtained by differentiating a horizontal accelerogram and dividing by the shear wave speed for hard and soft soil conditions.

  12. Preliminary safety analysis for key design features of KALIMER

    International Nuclear Information System (INIS)

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions

  13. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    International Nuclear Information System (INIS)

    The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category 3. The final hazard categorization for the deactivated 340 Waste Handling Facility (340 Facility) is presented in this document. This hazard categorization was prepared in accordance with DOE-STD-1 027-92, Change Notice 1, Hazard Categorization and Accident Analysis Techniques for Compliance with Doe Order 5480.23, Nuclear Safety Analysis Reports. The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category (HC) 3. Routine nuclear waste receiving, storage, handling, and shipping operations at the 340 Facility have been deactivated, however, the facility contains a small amount of radioactive liquid and/or dry saltcake in two underground vault tanks. A seismic event and hydrogen deflagration were selected as bounding accidents. The generation of hydrogen in the vault tanks without active ventilation was determined to achieve a steady state volume of 0.33%, which is significantly less than the lower flammability limit of 4%. Therefore, a hydrogen deflagration is not possible in these tanks. The unmitigated release from a seismic event was used to categorize the facility consistent with the process defined in Nuclear Safety Technical Position (NSTP) 2002-2. The final sum-of-fractions calculation concluded that the facility is less than HC 3. The analysis did not identify any required engineered controls or design features. The Administrative Controls that were derived from the analysis are: (1) radiological inventory control, (2) facility change control, and (3) Safety Management Programs (SMPs). The facility configuration and radiological inventory shall be controlled to ensure that the assumptions in the analysis remain valid. The facility commitment to SMPs protects the integrity of the facility and environment by ensuring training, emergency response, and radiation protection. The full scale

  14. Tritium Research Laboratory safety analysis report

    International Nuclear Information System (INIS)

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment

  15. Tritium Research Laboratory safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Wright, D.A.

    1979-03-01

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment. (ERB)

  16. Thermohydraulic design and safety analysis of research reactors

    International Nuclear Information System (INIS)

    This contribution presents briefly the trend of thermal hydraulic design and safety analysis of medium and high flux research reactors. This field of deterministic safety analysis is being considered by the IAEA in the framework of coordinated research project (CRP) initiated in 2002 on the Assessment of Analytical Tools for Different Research Reactor Types. The objective of this project is to establish a forum of international experts in order to integrate the activities for improvement and verification of selected computer codes that can be considered as reference tools in the safety analysis of research reactors, similar to that by power reactors. This undertaken supports the international ambition in improving the safety features and standards of research reactors, which can be useful for countries with long experience on RR and very helpful for countries having research reactors with low neutron flux and may looking for to extend them or build other reactors with higher neutron flux. In this regard the methodological approach on modification, verification and application of advanced computer codes for the safety analysis of research reactors is presented. In this regard the methodological approach on modification, verification and application of advanced computer codes for the safety analysis of research reactors is presented. The presented paper deals with the prediction of a semi empirical correlation for the first design limit regarding the onset of flow instability. Using the experimental data by considering the verification results for the thermalhydraulic and safety analysis code ATHLET a simple correlation for the Onset of Flow Instability (OFI) for medium and high flux reactors has been suggested. This correlation predicts by a given maximum heat flux of the hot channel the amount of inlet flow velocity at which an onset of instability is expected. It presents a simple procedure to estimate the minimum allowed flow velocity at which the fuel element

  17. Accident consequence calculations for project W-058 safety analysis

    International Nuclear Information System (INIS)

    This document describes the calculations performed to determine the accident consequences for the W-058 safety analysis. Project W-058 is the replacement cross site transfer system (RCSTS), which is designed to transort liquid waste between the 200 W and 200 E areas. Calculations for RCSTS safety analyses used the same methods as the calculations for the Tank Waste Remediation System (TWRS) Basis for Interim Operation (BIO) and its supporting calculation notes. Revised analyses were performed for the spray and pool leak accidents since the RCSTS flows and pressures differ from those assumed in the TWRS BIO. Revision 1 of the document incorporates review comments

  18. Statistical analysis applied to safety culture self-assessment

    International Nuclear Information System (INIS)

    Interviews and opinion surveys are instruments used to assess the safety culture in an organization as part of the Safety Culture Enhancement Programme. Specific statistical tools are used to analyse the survey results. This paper presents an example of an opinion survey with the corresponding application of the statistical analysis and the conclusions obtained. Survey validation, Frequency statistics, Kolmogorov-Smirnov non-parametric test, Student (T-test) and ANOVA means comparison tests and LSD post-hoc multiple comparison test, are discussed. (author)

  19. Fuel Storage Facility Final Safety Analysis Report. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Linderoth, C.E.

    1984-03-01

    The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.

  20. PBMR nuclear design and safety analysis: An overview

    International Nuclear Information System (INIS)

    PBMR is a high-temperature helium-cooled graphite-moderated continuous-fuelled pebble bed reactor. The power conversion unit is directly coupled to the reactor and the power turbines are driven through a direct closed-circuit helium cycle. An overview is presented on the nuclear engineering analyses used for the design and safety assessment for the PBMR. Topics addressed are the PBMR design, safety and licensing requirements, nuclear engineering analysis results, software verification and validation, and advances in software development. (authors)

  1. Safety analysis report for packaging (onsite) multicanister overpack cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  2. The application of new mathematical structures to safety analysis

    International Nuclear Information System (INIS)

    Probabilistic safety analyses (PSAs) often depend on significant subjectivity. The recent successes of fuzzy logic and fuzzy and hybrid mathematics in portraying subjectivity is a reminder that a selection made from the most applicable mathematical tools is more important than forced adaptation of conventional tools. In this paper, the authors consider new approaches that enhance conventional and fuzzy PSA by improved handling of subjectivity. The most significant of the mathematical structures were have investigated (from a standpoint of safety analysis applications) will be described, and the general types of applications will be outlined

  3. Safety analysis report for packaging (onsite) multicanister overpack cask

    International Nuclear Information System (INIS)

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area

  4. Environmental and safety envelope analysis for inertial fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Freiwald, J.G.; Pendergrass, J.H.; Frank, T.G.

    1980-01-01

    This paper describes an envelope analysis concept and a generic process flow model which together can be used to identify and isolate plant functions and provide for detailed mass- and energy-balance bookkeeping for environmental and safety studies. Los Alamos Scientific Laboratory's (LASL) two laser fusion power plant concepts were analyzed with this approach. Samples of the detailed tables of material flow rates into and out of an envelope are presented in this paper. The tritium and lithium inventories and air activation were identified as having important potential environmental problems and safety risks.

  5. QuantUM: Quantitative Safety Analysis of UML Models

    CERN Document Server

    Leitner-Fischer, Florian; 10.4204/EPTCS.57.2

    2011-01-01

    When developing a safety-critical system it is essential to obtain an assessment of different design alternatives. In particular, an early safety assessment of the architectural design of a system is desirable. In spite of the plethora of available formal quantitative analysis methods it is still difficult for software and system architects to integrate these techniques into their every day work. This is mainly due to the lack of methods that can be directly applied to architecture level models, for instance given as UML diagrams. Also, it is necessary that the description methods used do not require a profound knowledge of formal methods. Our approach bridges this gap and improves the integration of quantitative safety analysis methods into the development process. All inputs of the analysis are specified at the level of a UML model. This model is then automatically translated into the analysis model, and the results of the analysis are consequently represented on the level of the UML model. Thus the analysi...

  6. Worker Safety and Health and Nuclear Safety Quarterly Performance Analysis (January - March 2008)

    Energy Technology Data Exchange (ETDEWEB)

    Kerr, C E

    2009-10-07

    The DOE Office of Enforcement expects LLNL to 'implement comprehensive management and independent assessments that are effective in identifying deficiencies and broader problems in safety and security programs, as well as opportunities for continuous improvement within the organization' and to 'regularly perform assessments to evaluate implementation of the contractor's processes for screening and internal reporting.' LLNL has a self-assessment program, described in ES&H Manual Document 4.1, that includes line, management and independent assessments. LLNL also has in place a process to identify and report deficiencies of nuclear, worker safety and health and security requirements. In addition, the DOE Office of Enforcement expects LLNL to evaluate 'issues management databases to identify adverse trends, dominant problem areas, and potential repetitive events or conditions' (page 14, DOE Enforcement Process Overview, December 2007). LLNL requires that all worker safety and health and nuclear safety noncompliances be tracked as 'deficiencies' in the LLNL Issues Tracking System (ITS). Data from the ITS are analyzed for worker safety and health (WSH) and nuclear safety noncompliances that may meet the threshold for reporting to the DOE Noncompliance Tracking System (NTS). This report meets the expectations defined by the DOE Office of Enforcement to review the assessments conducted by LLNL, analyze the issues and noncompliances found in these assessments, and evaluate the data in the ITS database to identify adverse trends, dominant problem areas, and potential repetitive events or conditions. The report attempts to answer three questions: (1) Is LLNL evaluating its programs and state of compliance? (2) What is LLNL finding? (3) Is LLNL appropriately managing what it finds? The analysis in this report focuses on data from the first quarter of 2008 (January through March). This quarter is analyzed within the context of

  7. Computer analysis of thermal hydraulics for nuclear reactor safety

    International Nuclear Information System (INIS)

    This paper gives an overview of ANSTO's capability and recent research and development activities in thermal hydraulic modelling for nuclear reactor safety analysis, particularly for our research reactor, HIFAR (High Flux Australian Reactor) and its intended replacement, the Replacement Research Reactor (RRR). Several tools contribute to ANSTO's capability in thermal hydraulic modelling, including RELAP (developed in US) - a code for reactor system thermal-hydraulic analysis; CFS (developed in UK) - a general computational fluid dynamics code , which was used for thermal hydraulic analysis in reactor fuel elements; and HIZAPP (developed at ANSTO) - for coupling neutronics with thermal-hydraulics for reactor transient analysis

  8. Methods of probabilistic safety analysis for nuclear power plants - December 1996

    International Nuclear Information System (INIS)

    By identifying essential knowledge gaps, probabilistic safety analyses (PSA) provide insights which may affect the setting of priorities in reactor safety research including both plant safety and reduction of analysis insecurities. PSA is suitable for revealing methods and assumptions of safety assessment. No safety assessment of nuclear power plants (e.g. German Risk Study phase A and B) should do without is. (DG)

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

    International Nuclear Information System (INIS)

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

  10. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

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

  11. Using of BEPU methodology in a final safety analysis report

    International Nuclear Information System (INIS)

    The Nuclear Reactor Safety (NRS) has been established since the discovery of nuclear fission, and the occurrence of accidents in Nuclear Power Plants worldwide has contributed for its improvement. The Final Safety Analysis Report (FSAR) must contain complete information concerning safety of the plant and plant site, and must be seen as a compendium of NRS. The FSAR integrates both the licensing requirements and the analytical techniques. The analytical techniques can be applied by using a realistic approach, addressing the uncertainties of the results. This work aims to show an overview of the main analytical techniques that can be applied with a Best Estimated Plus Uncertainty (BEPU) methodology, which is 'the best one can do', as well as the ALARA (As Low As Reasonably Achievable) principle. Moreover, the paper intends to demonstrate the background of the licensing process through the main licensing requirements. (author)

  12. Computational Methods for Sensitivity and Uncertainty Analysis in Criticality Safety

    International Nuclear Information System (INIS)

    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

  13. A proposal for performing software safety hazard analysis

    International Nuclear Information System (INIS)

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

  14. A proposal for performing software safety hazard analysis

    International Nuclear Information System (INIS)

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

  15. Unavailability Analysis of Digital Engineered Safety Feature Actuation System

    International Nuclear Information System (INIS)

    This paper quantitatively presents the results of the fault tree analysis of Digital Engineered Safety Feature Actuation System which is one of the most important signal generation systems in nuclear power plant because it generates the signal for mitigating possible accidents. In this paper, as an example, we explore the case of auxiliary feedwater actuation signal. Based on the analysis results, we quantitatively explain the relationship between the important characteristics of digital systems and the system unavailability. We find out some factors which remarkably affect the system unavailability. They are the common cause failures and the coverage of fault tolerant mechanisms. Human operator's backup also plays very important role. In this analysis we ignore the effect of software failure. We also compare the result with the PSA result of conventional analog Engineered Safety Feature Actuation System. The result of Digital ESFAS is about 27% lower than in the analog system

  16. Performance and safety analysis of WP-cave concept

    International Nuclear Information System (INIS)

    The report presents a performance safety, and cost analysis of the WP-cave, WPC, concept. In the performance analysis, questions specific to the WPC have been addressed which have been identified to require more detailed studies. Based on the outcome of this analysis, a safety analysis has been made which comprises of the modeling and calculation of radionuclide transport from the repository to the biosphere and the resulting dose exposure to man. The result of the safety analysis indicates that the present design of a WPC repository may give unacceptably high doses. By improving the properties of the bentonite/sand barrier such that the hydraulic conductivity is reduced, or by changing the short-lived steel canisters to more long-lived canisters, e.g. copper canisters, it is judged possible to achieve a sufficiently low level of dose exposure rates to man. The cost for a WPC repository of the studied design is significantly higher than for a KBS-3 repository considering the Swedish conditions and the Swedish amount of spent fuel. The major costs are connected to the excavation and backfilling of the bentonite/sand barrier. The potential for cost savings is high but it is not judged possible to account for savings in such a way that the WPC concept shows lower cost than the KBS-3 concept. (34 figs., 33 tabs., 29 refs.)

  17. Current status of methodologies for seismic probabilistic safety analysis

    International Nuclear Information System (INIS)

    This report is a review of the methodology for conducting a seismic-probabilistic safety analysis (PSA) at a nuclear power station. The objective of this review is to provide an up-to-date review of the state-of-the-art of the various sub-methodologies that comprise the overall seismic-PSA methodology for addressing the safety of nuclear power stations, plus an overview of the whole methodological picture. In preparing this review, the author has had in mind several categories of readers and users: policy-level decision-makers (such as managers of nuclear power stations and regulators of nuclear safety), seismic-PSA practitioners, and PSA practitioners more broadly. The review concentrates on evaluating the extent to which today's seismic-PSA methodology produces reliable and useful results and insights, at its current state-of-the-art level, for assessing nuclear-power-station safety. Also, this review paper deals exclusively with seismic-PSA for addressing nuclear-power-station safety. Because the author is based in the U.S., it is natural that this review will contain more emphasis on U.S. experience than on experience in other countries. However, significant experience elsewhere is a major part of the basis for this evaluation

  18. A new database for food safety: EDID (Endocrine disrupting chemicals Diet Interaction Database)

    International Nuclear Information System (INIS)

    Diet is a significant source of exposure to endocrine disrupting chemicals (EDC); health risks cannot be excluded, in particular long-term effects in vulnerable groups such as children. However, food safety assessment must also consider the effects of natural food components modulating the endocrine system. The scientific evidence on the complex interactions between EDC and food components is still limited. The new EDC-Diet Interactions Database (EDID) within the ISS EDC area (www.iss.it/inte/) aims to stimulate further research in the field of food toxicology: a database on international literature's studies, either on experimental systems and on animal population and humans, easy to consult and periodically updated. Examples of studies contained in EDID are provided concerning EDC with iodine, vitamins and phyto estrogens

  19. Applicability of trends in nuclear safety analysis to space nuclear power systems

    International Nuclear Information System (INIS)

    A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication

  20. Safety evaluation review of the prototype license application safety analysis report

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the belowground vault (BGV) alternative method of low-level radioactive waste disposal. In Volume 1 of NUREG-1375, the NRC staff provided the safety review results for an earth-mounded concrete bunker PLASAR. In the current report, the staff focused its review on the design, construction, and operational aspects of the BGV PLASAR. The staff developed review comments and questions using the Standard Review Plan (SRP), Rev. 1 (NUREG-1200) as the basis for evaluating the acceptability of the information provided in the BGV PLASAR. The detailed review comments provided in this report are intended to be useful guidance to facility developers and State regulators in addressing issues likely to be encountered in the review of a license application for a low-level-waste disposal facility. 44 refs

  1. ‘Geo’chemical research: A key building block for nuclear waste disposal safety cases

    Science.gov (United States)

    Altmann, Scott

    2008-12-01

    Disposal of high level radioactive waste in deep underground repositories has been chosen as solution by several countries. Because of the special status this type waste has in the public mind, national implementation programs typically mobilize massive R&D efforts, last decades and are subject to extremely detailed and critical social-political scrutiny. The culminating argument of each program is a 'Safety Case' for a specific disposal concept containing, among other elements, the results of performance assessment simulations whose object is to model the release of radionuclides to the biosphere. Public and political confidence in performance assessment results (which generally show that radionuclide release will always be at acceptable levels) is based on their confidence in the quality of the scientific understanding in the processes included in the performance assessment model, in particular those governing radionuclide speciation and mass transport in the geological host formation. Geochemistry constitutes a core area of research in this regard. Clay-mineral rich formations are the subjects of advanced radwaste programs in several countries (France, Belgium, Switzerland…), principally because of their very low permeabilities and demonstrated capacities to retard by sorption most radionuclides. Among the key processes which must be represented in performance assessment models are (i) radioelement speciation (redox state, speciation, reactions determining radionuclide solid-solution partitioning) and (ii) diffusion-driven transport. The safety case must therefore demonstrate a detailed understanding of the physical-chemical phenomena governing the effects of these two aspects, for each radionuclide, within the geological barrier system. A wide range of coordinated (and internationally collaborated) research has been, and is being, carried out in order to gain the detailed scientific understanding needed for constructing those parts of the Safety Case

  2. Safety

    International Nuclear Information System (INIS)

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

  3. Transition towards replacing animal tests in safety assessment of cosmetics and chemicals: a combined TIS-MLP framework

    OpenAIRE

    Kooijman, M.; van der Meer, P.; Moors, E.H.M.; Schellekens, H; Hekkert, M.P.

    2012-01-01

    The urgency of the transition to replace animal tests in safety assessment of chemicals and cosmetics was triggered by societal resistance to animal testing (Rowan, 2007) and the scientific dispute concerning the value of animal testing (Olson et al., 2000). Since the 1980s the European Union (EU) has been developing policies to reduce an-imal studies. However, these policies have not been very successful, since only a few regulatory safety assessments in animals (among which the Draize eye t...

  4. Advances In Burnup Credit Criticality Safety Analysis Methods And Applications

    International Nuclear Information System (INIS)

    An International Workshop on “Advances in Applications of Burnup Credit for Spent Fuel Storage, Transport, Reprocessing, and Disposition” organized by the Nuclear Safety Council of Spain (CSN) in cooperation with the International Atomic Energy Agency (IAEA) was held at Córdoba, Spain, on October 27– 30, 2009. The objectives of this workshop were to identify the benefits that accrue from recent improvements of the burnup credit (BUC) analysis methodologies, to analyze the implications of applying improved BUC methodologies, focusing on both the safety-related and operational aspects, and to foster the exchange of international experience in licensing and implementation of BUC applications. In the paper on hand the attention is focused on the improvements of BUC analysis methodologies. (author)

  5. Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

    International Nuclear Information System (INIS)

    This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on four type B Packages: the 9972, 9973, 9974, and 9975 packages. Because all four packages have similar designs with very similar performance characteristics, all of them are presented in a single SARP. The performance evaluation presented in this SARP documents the compliance of the 9975 package with the regulatory safety requirements. Evaluations of the 9972, 9973, and 9974 packages support that of the 9975. To avoid confusion arising from the inclusion of four packages in a single document, the text segregates the data for each package in such a way that the reader interested in only one package can progress from Chapter 1 through Chapter 9. The directory at the beginning of each chapter identifies each section that should be read for a given package. Sections marked ''all'' are generic to all packages

  6. Methods and criteria for safety analysis (FIN L2535)

    International Nuclear Information System (INIS)

    In response to the NRC request for a proposal dated October 20, 1992, Westinghouse Savannah River Company (WSRC) submit this proposal to provide contractural assistance for FIN L2535, ''Methods and Criteria for Safety Analysis,'' as specified in the Statement of Work attached to the request for proposal. The Statement of Work involves development of safety analysis guidance for NRC licensees, arranging a workshop on this guidance, and revising NRC Regulatory Guide 3.52. This response to the request for proposal offers for consideration the following advantages of WSRC in performing this work: Experience, Qualification of Personnel and Resource Commitment, Technical and Organizational Approach, Mobilization Plan, Key Personnel and Resumes. In addition, attached are the following items required by the NRC: Schedule II, Savannah River Site - Job Cost Estimate, NRC Form 189, Project and Budget Proposal for NRC Work, page 1, NRC Form 189, Project and Budget Proposal for NRC Work, page 2, Project Description

  7. Code conversion for system design and safety analysis of NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)

  8. Reliability and safety analysis for systems of fusion device

    Energy Technology Data Exchange (ETDEWEB)

    Alzbutas, Robertas, E-mail: robertas.alzbutas@lei.lt; Voronov, Roman

    2015-05-15

    Highlights: • Reliability is very important from fusion devices efficiency perspective. • Rich experience of probabilistic safety analysis exists in nuclear industry. • Reliability and safety analysis was applied for systems of fusion device. • This enables to identify and prioritize availability improvement measures. • Recommendations are based on cost effectiveness for risk decrease options. - Abstract: Fusion energy or thermonuclear power is a promising, literally endless source of energy. Development of fusion power is still under investigation and experimental phase, and a number of fusion devices are under construction in Europe. Since fusion energy is innovative and fusion devices contain unique and expensive equipment, an issue of their reliability is very important from their efficiency perspective. A Reliability, Availability, Maintainability, Inspectability (RAMI) analysis is being performed or is going to be performed in the nearest future for such fusion devices as ITER and DEMO in order to ensure reliable and efficient operation for experiments (e.g., in ITER) or for energy production purposes (e.g., in DEMO). On the other hand, rich experience of the reliability and Probabilistic Safety Analysis (PSA) exists in nuclear industry for fission power plants and other nuclear installations. In this paper, the Wendelstein 7-X (W7-X) device is mainly considered. This stellarator device is in commissioning stage in the Max-Planck-Institut für Plasmaphysik, Greifswald, Germany (IPP). In the frame of cooperation between the IPP and the Lithuanian Energy Institute (LEI) under the European Fusion Development Agreement a pilot project of a reliability analysis of the W7-X systems was performed with a purpose to adopt Nuclear Power Plant (NPP) PSA experience for fusion device systems. During the project reliability and safety (risk) analysis of a Divertor Target Cooling Circuit, which is an important system for permanent and reliable operation of in

  9. Reliability and safety analysis for systems of fusion device

    International Nuclear Information System (INIS)

    Highlights: • Reliability is very important from fusion devices efficiency perspective. • Rich experience of probabilistic safety analysis exists in nuclear industry. • Reliability and safety analysis was applied for systems of fusion device. • This enables to identify and prioritize availability improvement measures. • Recommendations are based on cost effectiveness for risk decrease options. - Abstract: Fusion energy or thermonuclear power is a promising, literally endless source of energy. Development of fusion power is still under investigation and experimental phase, and a number of fusion devices are under construction in Europe. Since fusion energy is innovative and fusion devices contain unique and expensive equipment, an issue of their reliability is very important from their efficiency perspective. A Reliability, Availability, Maintainability, Inspectability (RAMI) analysis is being performed or is going to be performed in the nearest future for such fusion devices as ITER and DEMO in order to ensure reliable and efficient operation for experiments (e.g., in ITER) or for energy production purposes (e.g., in DEMO). On the other hand, rich experience of the reliability and Probabilistic Safety Analysis (PSA) exists in nuclear industry for fission power plants and other nuclear installations. In this paper, the Wendelstein 7-X (W7-X) device is mainly considered. This stellarator device is in commissioning stage in the Max-Planck-Institut für Plasmaphysik, Greifswald, Germany (IPP). In the frame of cooperation between the IPP and the Lithuanian Energy Institute (LEI) under the European Fusion Development Agreement a pilot project of a reliability analysis of the W7-X systems was performed with a purpose to adopt Nuclear Power Plant (NPP) PSA experience for fusion device systems. During the project reliability and safety (risk) analysis of a Divertor Target Cooling Circuit, which is an important system for permanent and reliable operation of in

  10. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    International Nuclear Information System (INIS)

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

  11. Reactor safety analysis computer program features that enhance user productivity

    International Nuclear Information System (INIS)

    This paper describes several design features of the MARY computer program that increase user productivity. The MARY program was used to analyze behavior of the Savannah River Site (SRS) K Reactor during postulated nuclear and thermal-hydraulic transients, such as overpower and underflow events, before K Reactor was placed in cold standby in 1993. These analyses provide the bases for portions of the accident chapter of the K-Reactor Safety Analysis Report

  12. Criticality safety and shielding analysis of WWER-440 fuel configurations

    International Nuclear Information System (INIS)

    An overview is made of some studies performed on the criticality safety and radiation shielding analysis of irradiated WWER-440 fuel storage and handling configurations. The analytical tools are based on the SCALE 4.4a code system, in combination with the TORT discrete ordinates transport code and the BUGLE-96 cross-sections library. The accuracy of some important results is assessed through comparison with independent evaluations and with measurement data. (author)

  13. Modeling of Safety Functions in Quantitative Risk Analysis

    OpenAIRE

    Nguyen, Thien Duy

    2012-01-01

    Quantitative risk analysis in the offshore industry is mandated by the Norwegian legislation. A literature survey is carried out, related to the current legislation from the Norwegian Petroleum Safety Authority (PSA) and supporting NORSOK standards. Process accidents on offshore installations, operating on the Norwegian continental shelf are emphasized. A risk picture is the synthesis of a risk assessment, describing the risk level. Requirements to the risk picture are discussed, and associat...

  14. Safety analysis report for packaging (onsite) sample pig transport system

    International Nuclear Information System (INIS)

    This Safety Analysis Report for Packaging (SARP) provides a technical evaluation of the Sample Pig Transport System as compared to the requirements of the U.S. Department of Energy, Richland Operations Office (RL) Order 5480.1, Change 1, Chapter III. The evaluation concludes that the package is acceptable for the onsite transport of Type B, fissile excepted radioactive materials when used in accordance with this document

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

    International Nuclear Information System (INIS)

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

  16. Safety analysis report for packaging (onsite) sample pig transport system

    Energy Technology Data Exchange (ETDEWEB)

    MCCOY, J.C.

    1999-03-16

    This Safety Analysis Report for Packaging (SARP) provides a technical evaluation of the Sample Pig Transport System as compared to the requirements of the U.S. Department of Energy, Richland Operations Office (RL) Order 5480.1, Change 1, Chapter III. The evaluation concludes that the package is acceptable for the onsite transport of Type B, fissile excepted radioactive materials when used in accordance with this document.

  17. The probabilistic safety analysis of Jose Cabrera NPP in the Context of the Periodic safety review

    International Nuclear Information System (INIS)

    In July 1989, the Spanish Nuclear Safety Council (CSN) called on Jose Cabrera NPP (JCNPP) to perform a probabilistic safety analysis (PSA). Edition 1 of this PSA was presented in July 1993. Edition 2 was delivered to the CSN, along with the database of items pending from the evaluation of Edition 1, December 1997. In October 1998, the CSN and JCNPP agreed on the appropriateness of having a PSA approved for use in the evaluation of the Periodic Safety Review (PSR) and in the renewal process of the Provisional Operating Permit (October 1999). This involved a great effort on the part of both parties, who established a joint calendar of actions to be taken, setting strict deadlines. The deadline for delivering Edition 3 (models, data and quantification programmes was set for 15 june 1999. This was complemented by the preparation of applications on licensing-related issues, and a document reflecting the resolution of pending items. Subsequently, In April, JCNPP was required to prepare additional applications. (Author)

  18. Chemical management and control strategies: experiences from the GTZ pilot project on chemical safety in Indonesian small and medium-sized enterprises.

    Science.gov (United States)

    Tischer, M; Scholaen, S

    2003-10-01

    In 1998 the Deutsche Gesellschaft für Technische Zusammenarbeit (GTZ) launched the Convention Project on Chemical Safety in developing countries. The project aims to support developing countries in the implementation of the Rotterdam and Stockholm Conventions, create human resources and institutional capacities and to demonstrate via pilot measures how chemical safety in the partner countries can be improved and sustainably implemented in line with international standards. With this objective the development of a Chemical Management Guide (CM Guide) for small and medium-sized enterprises in developing countries has been initiated. The guide describes a step-by-step approach which is based on identifying 'hot-spots' as a first step, and making a chemical inventory as a second step. The third step is the continuous improvement of chemical management. In total, there are six tools that aim to support the chemical management process: basic concepts for risk assessment; description of control approaches; using material safety data sheets (MSDSs); risk phrases for hazardous substances; safety phrases for hazardous substances; symbols used for labelling hazardous substances. In the course of the test-implementation of the CM Guide in Indonesia, it was found that MSDSs were not available in most of the smaller companies. In contrast, medium-sized and larger companies do have more MSDSs available. It was also found that the way to engage the minds of company owners and managers is with economic arguments related to the loss, waste and expiry of materials, and quality standards expected from importing countries. PMID:14530183

  19. Development and Assessment of Best Estimate Integrated Safety Analysis Code

    International Nuclear Information System (INIS)

    The integrated safety analysis code MARS3.0 has been developed and assessed through v and v procedure. Integrated safety analysis system has been established through coupling with severe accident code and utilizing MARS subchannel capability. The coupled containment module has been also improved. Development of indigenous thermal hydraulic models for MARS3.0 has been done through the implementation of multidimensional two phase flow model, APR1400, SMART safety issue models and new reactor models. Development of droplet field model has been also attempted and implemented to trial version. The full scope assessment has been carried out for the system analysis module and 3D vessel module. The code has been also assessed through participating international cooperation programs. The experimental data needed to code assessment has been collected and maintained through the WEB based data bank program. 3D GUI(graphic user interface) has been developed for MARS users. MARS users group has been organized, and currently it consists of 22 domestic organizations, including research, industrial, regulatory organizations and universities

  20. Thermal hydraulic and safety analysis for Tajoura Research Center

    International Nuclear Information System (INIS)

    Thermal hydraulic and safety analysis of Tajoura Research Center (TRR) utilizing low enriched uranium (LEU) fuel type IRT-4M have been performed using computer code PARET. The compact loading of the present core comprises of 16 fuel assemblies and 11 control rods. Results of the thermal hydraulic analysis show that the reactor can be operated at steady-state power of 10 MW for a flow rate of 533 m3/h, with sufficient margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions, the considered events are, positive reactivity insertion, flow reduction due to loss of primary coolant. For each of these transients, time history of reactor power, energy releases, clad surface and fuel centerline temperatures and maximum heat flux ratios were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these transients. It is therefore concluded that the reactor can be operated at steady-state power level of 10 MW without compromising safety. (author)

  1. Development and assessment of best estimate integrated safety analysis code

    International Nuclear Information System (INIS)

    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

  2. Safety analysis of the existing 804 and 845 firing facilities

    International Nuclear Information System (INIS)

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 804 and 845 Firing Facilities at Site 300 could present undue hazards to the general public, peronnel at Site 300, or have an adverse effect on the environment. The normal operation and credible accident that might have an effect on these facilities or have off-site consequence were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives. Since this hazard has the potential for causing significant on-site and minimum off-site consequences, Bunkers 804 and 845 have been classified as moderate hazard facilties per DOE Order 5481.1A. This safety analysis concluded that the operation at these facilities will present no undue risk to the health and safety of LLNL employees or the public

  3. 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.

  4. Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

    International Nuclear Information System (INIS)

    This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

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

    International Nuclear Information System (INIS)

    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

  6. Safety Analysis for Sub-channel Blockage in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jin; Chang, Wonpyo; Ha, Kisuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The flow perturbation caused by the blockage could raise the local coolant temperature in the incident and it might eventually lead to the degradation of the fuel rods. Therefore, a partial flow blockage accident must be a safety concern in the SFR design. In this regard, analyses were performed for the flow blockage accident postulated in a conceptual design of a 150MWe Proto-type SFR using the MATRA-LMR/FB and analysis result was compared to the safety acceptance criterion shown in Table 1 developed by KAERI. The maximum coolant temperatures for 6, 24 channels blockage occurred at the end of the fuel slug and both of them satisfied the safety limits. However, for the 54 channels blockage, the maximum coolant temperature was found in the downstream of the blockage and it could not meet the safety limits. It was caused by the recirculation region in the downstream of the blockage. In conclusion, satisfactory margins were obtained for 6, 24 channel blockage cases.

  7. Reliability and safety analysis of redundant vehicle management computer system

    Institute of Scientific and Technical Information of China (English)

    Shi Jian; Meng Yixuan; Wang Shaoping; Bian Mengmeng; Yan Dungong

    2013-01-01

    Redundant techniques are widely adopted in vehicle management computer (VMC) to ensure that VMC has high reliability and safety. At the same time, it makes VMC have special char-acteristics, e.g., failure correlation, event simultaneity, and failure self-recovery. Accordingly, the reliability and safety analysis to redundant VMC system (RVMCS) becomes more difficult. Aimed at the difficulties in RVMCS reliability modeling, this paper adopts generalized stochastic Petri nets to establish the reliability and safety models of RVMCS. Then this paper analyzes RVMCS oper-ating states and potential threats to flight control system. It is verified by simulation that the reli-ability of VMC is not the product of hardware reliability and software reliability, and the interactions between hardware and software faults can reduce the real reliability of VMC obviously. Furthermore, the failure undetected states and false alarming states inevitably exist in RVMCS due to the influences of limited fault monitoring coverage and false alarming probability of fault mon-itoring devices (FMD). RVMCS operating in some failure undetected states will produce fatal threats to the safety of flight control system. RVMCS operating in some false alarming states will reduce utility of RVMCS obviously. The results abstracted in this paper can guide reliable VMC and efficient FMD designs. The methods adopted in this paper can also be used to analyze other intelligent systems’ reliability.

  8. Thermal analysis and safety information for metal nanopowders by DSC

    International Nuclear Information System (INIS)

    Highlights: • Metal nanopowders are common and frequently employed in industry. • Nano iron powder experimental results of To were 140–150 °C. • Safety information can benefit relevant metal powders industries. - Abstract: Metal nanopowders are common and frequently employed in industry. Iron is mostly applied in high-performance magnetic materials and pollutants treatment for groundwater. Zinc is widely used in brass, bronze, die casting metal, alloys, rubber, and paints, etc. Nonetheless, some disasters induced by metal powders are due to the lack of related safety information. In this study, we applied differential scanning calorimetry (DSC) and used thermal analysis software to evaluate the related thermal safety information, such as exothermic onset temperature (To), peak of temperature (Tp), and heat of reaction (ΔH). The nano iron powder experimental results of To were 140–150 °C, 148–158 °C, and 141–149 °C for 15 nm, 35 nm, and 65 nm, respectively. The ΔH was larger than 3900 J/g, 5000 J/g, and 3900 J/g for 15 nm, 35 nm, and 65 nm, respectively. Safety information can benefit the relevant metal powders industries for preventing accidents from occurring

  9. Application of Computer Integration Technology for Fire Safety Analysis

    Institute of Scientific and Technical Information of China (English)

    SHI Jianyong; LI Yinqing; CHEN Huchuan

    2008-01-01

    With the development of information technology, the fire safety assessment of whole structure or region based on the computer simulation has become a hot topic. However, traditionally, the concemed studies are performed separately for different objectives and difficult to perform an overall evaluation. A new multi-dimensional integration model and methodology for fire safety assessment were presented and two newly developed integrated systems were introduced to demonstrate the function of integration simulation technology in this paper. The first one is the analysis on the fire-resistant behaviors of whole structure under real fire loads. The second one is the study on fire evaluation and emergency rescue of campus based on geography information technology (GIS). Some practical examples are presented to illuminate the advan-tages of computer integration technology on fire safety assessment and emphasize some problems in the simulation. The results show that the multi-dimensional integration model offers a new way and platform for the integrating fire safety assessment of whole structure or region, and the integrated software developed is the useful engineering tools for cost-saving and safe design.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  12. 78 FR 73756 - Process Safety Management and Prevention of Major Chemical Accidents

    Science.gov (United States)

    2013-12-09

    ... Occupational Safety and Health Administration 29 CFR Part 1910 RIN 1218-AC82 Process Safety Management and... requests comment on potential revisions to its Process Safety Management (PSM) standard and its Explosives...://www.osha.gov/SLTC/processsafetymanagement/ . B. Process Safety Management of Highly...

  13. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    International Nuclear Information System (INIS)

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly

  14. Waste sampling and characterization facility complex safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Meloy, R.T., Westinghouse Hanford

    1996-06-04

    The Waste Sampling and Characterization Facility is a `Non-Nuclear, Radiological Facility. This document demonstrates, by analysis, that WSCF can meet the chemical and radiological inventory limits for a radiological facility. It establishes control that ensures those inventories are maintained below threshold values to preserve the `Non- Nuclear, Radiological` classification.

  15. Probabilistic analysis of safety in industrial irradiation plants

    International Nuclear Information System (INIS)

    The Argentinean Nuclear Regulatory Authority is carrying out the Probabilistic Safety Analysis (PSA) of the two industrial irradiation plants existent in the country. The objective of this presentation is to show from the regulatory point of view, the advantages of applying this tool, as well as the appeared difficulties; for it will be made a brief description of the facilities, of the method and of the normative one. Both plants are multipurpose facilities classified as 'industrial irradiator category IV' (panoramic irradiator with source deposited in pool). Basically, the execution of an APS consists of the following stages: 1. Identification of initiating events. 2. Modeling of Accidental Sequences (Event Trees). 3. Analysis of Systems (Fault trees). 4. Quantification of Accidental Sequences. The argentine normative doesn't demand to these facilities the realization of an APS, however the basic standard of Radiological Safety establishes that in the design of this type of facilities in the cases that is justified, should make sure that the annual probability of occurrence of an accidental sequence and the resulting dose in a person gives as result an radiological risk inferior to the risk limit adopted as acceptance criteria. On the other hand the design standard specifies for these irradiators it demands a maximum fault rate of 10-2 for the related components with the systems of radiological safety. In our case, the possible initiating events have been identified that carried out to not wanted situations (about people exposure, radioactive contamination). Then, for each one of the significant initiating events, the corresponding accidental sequences were modeled and the safety systems that intervene in this sequences by means of fault trees were analyzed, for then to determine the fault probabilities of the same ones. At the moment they are completing these fault trees, but the difficulty resides in the impossibility of obtaining real data of the reliability

  16. Classification analysis of organization factors related to system safety

    International Nuclear Information System (INIS)

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

  17. Management implementation plan for a safety analysis and review system

    International Nuclear Information System (INIS)

    The US Department of Energy has issued an Order, DOE 5481.1, which establishes uniform requirements for the preparation and review of Safety Analysis for DOE Operations. The Management Implementation Plan specified herein establishes the administrative procedures and technical requirements for implementing DOE 5481.1 to Operations under the cognizance of the Pittsburgh Energy Technology Center. This Implementation Plan is applicable to all present and future Operations under the cognizance of PETC. The Plan identifies those Operations for which DOE 5481.1 is applicable and those Operations for which no further analysis is required because the initial determination and review has concluded that DOE 5481.1 does not apply

  18. Deconvolution of variability and uncertainty in the Cassini safety analysis

    International Nuclear Information System (INIS)

    The standard method for propagation of uncertainty in a risk analysis requires rerunning the risk calculation numerous times with model parameters chosen from their uncertainty distributions. This was not practical for the Cassini nuclear safety analysis, due to the computationally intense nature of the risk calculation. A less computationally intense procedure was developed which requires only two calculations for each accident case. The first of these is the standard 'best-estimate' calculation. In the second calculation, variables and parameters change simultaneously. The mathematical technique of deconvolution is then used to separate out an uncertainty multiplier distribution, which can be used to calculate distribution functions at various levels of confidence

  19. Application of Fisher Discriminant Analysis in Safety Evaluation

    Directory of Open Access Journals (Sweden)

    ZHU XiaoZhen

    2016-06-01

    Full Text Available The multivariate statistical method of Fisher discriminant analysis is applied to safety evaluation, through the analysis of the original data, the assessment process, built up to reflect the evaluated object security status of evaluation function model, so as to simplify the subsequent similar evaluation target workload. The two mine in south of a mining enterprise subordinate to the environmental conditions in six integrated index evaluation, comprehensive index function model is established, finally, the Fisher discrimination obtained results with Bayesian discriminant obtained results, the correctness of the model is verified that the model reliability is high, and simple and practical.

  20. Interactive Safety Analysis Framework of Autonomous Intelligent Vehicles

    Directory of Open Access Journals (Sweden)

    Cui You Xiang

    2016-01-01

    Full Text Available More than 100,000 people were killed and around 2.6 million injured in road accidents in the People’s Republic of China (PRC, that is four to eight times that of developed countries, equivalent to 6.2 mortality per 10 thousand vehicles—the highest rate in the world. There are more than 1,700 fatalities and 840,000 injuries yearly due to vehicle crashes off public highways. In this paper, we proposed a interactive safety situation and threat analysis framework based on driver behaviour and vehicle dynamics risk analysis based on ISO26262…

  1. Organizational Culture and Safety Performance in the Manufacturing Companies in Malaysia: A Conceptual Analysis

    OpenAIRE

    Ong Choon Hee; Lim Lee Ping

    2014-01-01

    The purpose of this paper is to provide a conceptual analysis of organizational culture and safety performance in the manufacturing companies in Malaysia. Our conceptual analysis suggests that manufacturing companies that adopt group culture or hierarchical culture are more likely to demonstrate safety compliance and safety participation. Manufacturing companies that adopt rational culture or developmental culture are less likely to demonstrate safety compliance and safety participation. Give...

  2. Disclosure of hydraulic fracturing fluid chemical additives: analysis of regulations.

    Science.gov (United States)

    Maule, Alexis L; Makey, Colleen M; Benson, Eugene B; Burrows, Isaac J; Scammell, Madeleine K

    2013-01-01

    Hydraulic fracturing is used to extract natural gas from shale formations. The process involves injecting into the ground fracturing fluids that contain thousands of gallons of chemical additives. Companies are not mandated by federal regulations to disclose the identities or quantities of chemicals used during hydraulic fracturing operations on private or public lands. States have begun to regulate hydraulic fracturing fluids by mandating chemical disclosure. These laws have shortcomings including nondisclosure of proprietary or "trade secret" mixtures, insufficient penalties for reporting inaccurate or incomplete information, and timelines that allow for after-the-fact reporting. These limitations leave lawmakers, regulators, public safety officers, and the public uninformed and ill-prepared to anticipate and respond to possible environmental and human health hazards associated with hydraulic fracturing fluids. We explore hydraulic fracturing exemptions from federal regulations, as well as current and future efforts to mandate chemical disclosure at the federal and state level. PMID:23552653

  3. Tank waste compositions and atmospheric dispersion coefficients for use in accelerated safety analysis consequence assessments. Revision 1

    International Nuclear Information System (INIS)

    This topical report contains technical support information used to determine accident consequences for the Tank Farms Accelerated Safety Analysis (ASA) Interim Chapter 3, Hazard and Accident Analysis: Potential for Releases and Required Mitigation and Prevention and the Tank Waste Remediation System (TWRS) environmental impact statement (EIS) accident consequence report. It does not determine accident consequences or describe specific accident scenarios, but instead provides generic information used to calculate radiological and toxic chemical consequences for postulated tank farms accident releases

  4. Guidance for preparation of safety analysis reports for nonreactor facilities and operations

    International Nuclear Information System (INIS)

    Department of Energy (DOE) Orders 5480.23, ''Nuclear Safety Analysis Reports,'' and 5481.1B, ''Safety Analysis and Review System'' require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports

  5. Preliminary Safety Analysis of Korea HCSB Test Blanket Module

    International Nuclear Information System (INIS)

    A Helium Cooled Solid Breeder (HCSB) blanket has been considered as one of the promising blanket for the fusion power demonstration plant. Therefore HCSB Test Blanket Module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Korea has developed the HCSB TBM with some features such as graphite reflector and simplified flow passage. The objective of this study was to evaluate the thermal and structural integrity of the HCSB TBM under the hypothetical accidental conditions such as cooling pipe break in TBM. The safety analysis was performed under conservative conditions based on the TBM design, which can be assumed by the similarity of the safety analysis of the ITER shielding blanket. Transient analysis model was used to calculate the temperature distribution for Loss of Coolant Accident (LOCA). Simplified analysis conditions were a) simultaneous plasma shutdown and LOCA b) LOCA and then after FW temperature reaches 1150 deg. plasma shutdown. Helium circuit behavior during the different LOCA scenarios was also evaluated. Finally the design modifications based on the analysis result and the related R-and-D of the HCSB blanket design for the application in a DEMO reactor were mentioned. (author)

  6. Empirical Analysis of Construction Safety Climate - A Study

    OpenAIRE

    S.V.S.RAJA PRASAD; K.P.REGHUNATH

    2010-01-01

    Safety in the construction industry has always been a major issue. Though much improvement in construction safety has been achieved, the industry still continues to lag behind most other industries with regard to safety. The safety climate of any organization consists of employee’s attitudes towards and perceptions of, health and safety behavior. Construction workers attitudes towards safety are influenced by their perceptions of risk, management, safety rulesand procedures. A measure of safe...

  7. Safety analysis, 200 Area, Savannah River Plant: Separations area operations

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, W.C.; Lee, R.; Allen, P.M.; Gouge, A.P.

    1991-07-01

    The nev HB-Line, located on the fifth and sixth levels of Building 221-H, is designed to replace the aging existing HB-Line production facility. The nev HB-Line consists of three separate facilities: the Scrap Recovery Facility, the Neptunium Oxide Facility, and the Plutonium Oxide Facility. There are three separate safety analyses for the nev HB-Line, one for each of the three facilities. These are issued as supplements to the 200-Area Safety Analysis (DPSTSA-200-10). These supplements are numbered as Sup 2A, Scrap Recovery Facility, Sup 2B, Neptunium Oxide Facility, Sup 2C, Plutonium Oxide Facility. The subject of this safety analysis, the, Plutonium Oxide Facility, will convert nitrate solutions of {sup 238}Pu to plutonium oxide (PuO{sub 2}) powder. All these new facilities incorporate improvements in: (1) engineered barriers to contain contamination, (2) barriers to minimize personnel exposure to airborne contamination, (3) shielding and remote operations to decrease radiation exposure, and (4) equipment and ventilation design to provide flexibility and improved process performance.

  8. Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident

    International Nuclear Information System (INIS)

    A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)

  9. TVO-92 safety analysis of spent fuel disposal

    International Nuclear Information System (INIS)

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

  10. TVO-92 safety analysis of spent fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H. [Technical Research Centre of Finland, Espoo (Finland). Nuclear Engineering Lab.

    1993-08-01

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites.

  11. Safety analysis, 200 Area, Savannah River Plant: Separations area operations

    International Nuclear Information System (INIS)

    The nev HB-Line, located on the fifth and sixth levels of Building 221-H, is designed to replace the aging existing HB-Line production facility. The nev HB-Line consists of three separate facilities: the Scrap Recovery Facility, the Neptunium Oxide Facility, and the Plutonium Oxide Facility. There are three separate safety analyses for the nev HB-Line, one for each of the three facilities. These are issued as supplements to the 200-Area Safety Analysis (DPSTSA-200-10). These supplements are numbered as Sup 2A, Scrap Recovery Facility, Sup 2B, Neptunium Oxide Facility, Sup 2C, Plutonium Oxide Facility. The subject of this safety analysis, the, Plutonium Oxide Facility, will convert nitrate solutions of 238Pu to plutonium oxide (PuO2) powder. All these new facilities incorporate improvements in: (1) engineered barriers to contain contamination, (2) barriers to minimize personnel exposure to airborne contamination, (3) shielding and remote operations to decrease radiation exposure, and (4) equipment and ventilation design to provide flexibility and improved process performance

  12. ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

    International Nuclear Information System (INIS)

    The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

  13. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  14. Safety analysis and code development for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Development effort of computer codes applicable to nuclear fuel cycle facilities for assisting the task of NISA has been carried out. The work consists of 1) verification of criticality safety analysis codes : MVP and SCALE, 2) studies on burn-up credit applied methods, 3) preparation of non-uniformity effect calculation for criticality safety, 4) development of the new convenient library for shielding calculation based on JENDL-3.3 nuclear data, 5) development of a numerical simulation code DYMPL for analyzing abnormal transients of PUREX processes, 6) radiation dose evaluation code development for reprocessing facilities, 7) updating the dose evaluation data for the probabilistic environmental assessment code MACCS2-JF by emergency scenario. (author)

  15. In tank processing safety analysis program summary report. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Radder, J.A.

    1994-11-01

    The purpose of this summary report is to present results from the safety analysis work that was performed in support of the ``Seismic Safety Issue Resolution Program Plan`` for the In-Tank Processing (ITP) Facility. Results from this effort include estimates of the consequences that postulated earthquakes might introduce. For beyond evaluation based earthquake (EBE) events, best estimate values (e.g., waste tank volumes) are used rather than bounding values to analyze the consequences of such events. This is consistent with the probabilistic approach outlined in Attachment C of the program plan. Planned follow-on work will also involve best estimates of probabilities for soil liquefaction and differential settlement. These probabilities will be combined in an accident progression event tree (APET) model that is used to provide estimates of risk for beyond EBE seismic events.

  16. Criticality safety analysis of WWER spent fuel facilities

    International Nuclear Information System (INIS)

    Criticality safety analysis of the WWER transport casks as well as of WWER wet spent fuel storage at the Kozloduy NPP site is performed under conservative assumptions. For criticality calculations the methodology based on the modular code system SCALE4.4 and MCNP code has been applied. The criticality parameters for WWER-440 and WWER-1000 transport casks, and for WWER wet spent fuel storage have been calculated by SCALE. The verification of the results has been carried out based on the comparison with the MCNP results. This comparison shows good coincidence of the results, calculated by SCALE and MCNP, in the uncertainty margins. The results presented lead to the conclusion that the criticality safety criteria Keff<0.95 for WWER transport casks and WWER wet spent fuel storage are satisfied quite well under conservative assumptions

  17. Event analysis for safety relevance rating of human performance

    International Nuclear Information System (INIS)

    The paper describes a project for establishing instruments allowing identification and evaluation of safety relevance of human action without the need for prior, detailed PSA. Criteria and auxiliary means such as tabulated information for rapid acquisition, analysis and evaluation of available technical and organisational/administrative information have been elaborated, and are applied for safety relevance rating of human factors. The method has been applied to evaluate human performance under specified normal operating conditions including in-service inspections. Human performance in notifiable events was analysed, and particularly comprehensive examinations have been carried out relating to maintenance and repair work, in-service inspections, standard switching processes, modifications, and work scheduling. (orig./CB)

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

    International Nuclear Information System (INIS)

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

  19. CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application

    International Nuclear Information System (INIS)

    Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems

  20. The development of technologies of safety analysis for LMR ('03)

    International Nuclear Information System (INIS)

    The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C up to the

  1. 10 CFR 52.79 - Contents of applications; technical information in final safety analysis report.

    Science.gov (United States)

    2010-01-01

    ...; technical information in final safety analysis report. (a) The application must contain a final safety... CFR part 50, appendix A, GDC 3, and § 50.48 of this chapter; (7) A description of protection provided... important to safety and the list of electric equipment important to safety that is required by 10 CFR...

  2. Current status of safety design and safety analysis for China ITER helium coolant ceramic breeder test blanket system long

    International Nuclear Information System (INIS)

    Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. (author)

  3. Hybrid chemical and nondestructive-analysis technique

    Energy Technology Data Exchange (ETDEWEB)

    Hsue, S.T.; Marsh, S.F.; Marks, T.

    1982-01-01

    A hybrid chemical/NDA technique has been applied at the Los Alamos National Laboratory to the assay of plutonium in ion-exchange effluents. Typical effluent solutions contain low concentrations of plutonium and high concentrations of americium. A simple trioctylphosphine oxide (TOPO) separation can remove 99.9% of the americium. The organic phase that contains the separated plutonium can be accurately assayed by monitoring the uranium L x-ray intensities.

  4. Hybrid chemical and nondestructive analysis technique

    International Nuclear Information System (INIS)

    A hybrid chemical/NDA technique has been applied at the Los Alamos National Laboratory to the assay of plutonium in ion-exchange effluents. Typical effluent solutions contain low concentrations of plutonium and high concentrations of americium. A simple trioctylphosphine oxide (TOPO) separation can remove 99.9% of the americium. The organic phase that contains the separated plutonium can be accurately assayed by monitoring the uranium L x-ray intensities

  5. Safety analysis and the code development on radioactive waste disposal

    International Nuclear Information System (INIS)

    In order to confirm the long-term safety concerning sub-surface disposal, we studied the function about the climatic and topographic changes included in three-dimensional groundwater flow analysis code 3D-SEEP. And, we studied the methods of the groundwater flow analysis and particle tracking analysis in consideration of long-term phenomenon. Moreover, we made the trial calculations of the long-term transient analysis using this function. As a result, we found the adaptation range of the code and the differences from the results obtained by the steady state analysis. As a reflection of new knowledge about the particle tracking analysis, we carried out the trial calculation which was adapted in the analysis technique in consideration of the geometry model which changes with time progress. As a result, we found the differences from the results obtained by the conventional method, and the present subjects. We introduced and improved the groundwater flow and nuclide migration analysis code MH-FLOW using the mixed hybrid finite element method. This analysis code was developed for the purpose of obtaining a solution with sufficient accuracy even for the heterogeneous place where coefficients of permeability is greatly different. Moreover, we used MH-FLOW for the benchmark problem defined in an international project, and compared results with those obtained by the project. As a result, we checked the validity of MH-FLOW. (author)

  6. Piping failure analysis in domestic nuclear safety piping system

    International Nuclear Information System (INIS)

    The purpose of this paper is to analyze piping failure trend of safety pipings in domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of pipping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in Chemical and Volume Control System (CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity of socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure cases

  7. Canister Storage Building (CSB) safety analysis report, phase 3: Safety analysis documentation supporting CSB construction

    International Nuclear Information System (INIS)

    The US Department of Energy established the K Basins Spent Nuclear Fuel Project to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored under water in the Hanford Site's K Basins, which are located near the Columbia River. Recommendations for a series of aggressive projects to construct and operate systems and facilities to manage the safe removal of K Basins fuel were made in WHC-EP-0830, Hanford Spent Nuclear Fuel Recommended Path Forward, and its subsequent update, WHC-SD-SNF-SP-005, Hanford Spent Nuclear Fuel Project Integrated Process Strategy for K Basins Fuel. The integrated process strategy recommendations include the following steps: Fuel preparation activities at the K Basins, including removing the fuel elements from their K Basin canisters, separating fuel particulate from fuel elements and fuel fragments greater than 0.6 cm (0.25 in.) in any dimension, removing excess sludge from the fuel and fuel fragments by means of flushing, as necessary, and packaging the fuel into multicanister overpacks (MCOs); Removal of free water by draining and vacuum drying at a cold vacuum drying facility ES-122; Dry shipment of fuel from the Cold Vacuum Drying to the Canister Storage Building (CSB), a new facility in the 200 East Area of the Hanford Site

  8. EDXRF for non-destructive chemical analysis

    International Nuclear Information System (INIS)

    One of the non-destructive methods used for the identification and verification of metals is by the energy-dispersive X-ray fluorescence (EDXRF) technique. EDXRF analysis provides several important advantages such as simultaneous determination of the elements present, enable to analyse a very wide concentration range, fast analysis with no sample preparation. The paper shows how this technique is developed and applied in the identification and verification of different grades of stainless steels and also precious metals analysis. (Author)

  9. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  10. Gap Analysis Approach for Construction Safety Program Improvement

    OpenAIRE

    Thanet Aksorn; B.H.W. Hadikusumo

    2007-01-01

    To improve construction site safety, emphasis has been placed on the implementation of safety programs. In order to successfully gain from safety programs, factors that affect their improvement need to be studied. Sixteen critical success factors of safety programs were identified from safety literature, and these were validated by safety experts. This study was undertaken by surveying 70 respondents from medium- and large-scale construction projects. It explored the importance and the actual...

  11. Analysis of Safety from a Human Clinical Trial with Pterostilbene

    Directory of Open Access Journals (Sweden)

    Daniel M. Riche

    2013-01-01

    Full Text Available Objectives. The purpose of this trial was to evaluate the safety of long-term pterostilbene administration in humans. Methodology. The trial was a prospective, randomized, double-blind placebo-controlled intervention trial enrolling patients with hypercholesterolemia (defined as a baseline total cholesterol ≥200 mg/dL and/or baseline low-density lipoprotein cholesterol ≥100 mg/dL. Eighty subjects were divided equally into one of four groups: (1 pterostilbene 125 mg twice daily, (2 pterostilbene 50 mg twice daily, (3 pterostilbene 50 mg + grape extract (GE 100 mg twice daily, and (4 matching placebo twice daily for 6–8 weeks. Safety markers included biochemical and subjective measures. Linear mixed models were used to estimate primary safety measure treatment effects. Results. The majority of patients completed the trial (91.3%. The average age was 54 years. The majority of patients were females (71% and Caucasians (70%. There were no adverse drug reactions (ADRs on hepatic, renal, or glucose markers based on biochemical analysis. There were no statistically significant self-reported or major ADRs. Conclusion. Pterostilbene is generally safe for use in humans up to 250 mg/day.

  12. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  13. Hydrogen safety : consequence analysis and hydrogen refuelling stations

    International Nuclear Information System (INIS)

    In addition to production and storage issues, the lack of specific standards for hydrogen as vehicle fuel is widely regarded as an obstacle to the introduction of hydrogen on the energy market. The tolerance of accidents involving emerging technologies, however rare, can be lower than for existing, familiar technologies, particularly when misperceptions of risk exist and little comparative risk analysis with more familiar fuels are available. This will result in a more conservative approach in establishing clearance distances and other safety issues for hydrogen refuelling stations by authorities having jurisdiction, which directly impact the cost of introducing a hydrogen refuelling infrastructure particularly in urban areas where real estate costs are high. Thus the lack of specific knowledge, experience and a clear and consistent methodology to assess the risk of hydrogen service stations and vehicles is widely regarded as a serious impediment to its use. In this presentation we present the R and D activities carried out under the Canadian Hydrogen Safety Programme and the Auto 21 Network of Centres of Excellence to develop a scientific and engineering basis for determining safety standards and safe industry practices specific to hydrogen, focusing on refuelling stations and hydrogen-fuelled vehicles. (author)

  14. Upgraded safety analysis document including operations policies, operational safety limits and policy changes. Revision 2

    International Nuclear Information System (INIS)

    The National Synchrotron Light Source Safety Analysis Reports (1), (2), (3), BNL reports number-sign 51584, number-sign 52205 and number-sign 52205 (addendum) describe the basic Environmental Safety and Health issues associated with the department's operations. They include the operating envelope for the Storage Rings and also the rest of the facility. These documents contain the operational limits as perceived prior or during construction of the facility, much of which still are appropriate for current operations. However, as the machine has matured, the experimental program has grown in size, requiring more supervision in that area. Also, machine studies have either verified or modified knowledge of beam loss modes and/or radiation loss patterns around the facility. This document is written to allow for these changes in procedure or standards resulting from their current mode of operation and shall be used in conjunction with the above reports. These changes have been reviewed by NSLS and BNL ES and H committee and approved by BNL management

  15. LOFT integral test system final safety analysis report

    International Nuclear Information System (INIS)

    Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)

  16. NUSAR: N Reactor Updated Safety Analysis Report, Amendment 21

    Energy Technology Data Exchange (ETDEWEB)

    Smith, G L

    1989-12-01

    The enclosed pages are Amendment 21 of the N Reactor Updated Safety Analysis Report (NUSAR). NUSAR, formerly UNI-M-90, was revised by 18 amendments that were issued by UNC Nuclear Industries, the contractor previously responsible for N Reactor operations. As of June 1987, Westinghouse Hanford Company (WHC) acquired the operations and engineering contract for N Reactor and other facilities at Hanford. The document number for NUSAR then became WHC-SP-0297. The first revision was issued by WHC as Amendment 19, prepared originally by UNC. Summaries of each of the amendments are included in NUSAR Section 1.1.

  17. Safety analysis for the 233-S decontamination and decommissioning project

    International Nuclear Information System (INIS)

    Decommissioning of the 233-S Plutonium Concentration Facility (REDOX) is a proposed expedited response action that is regulated by the Comprehensive Environmental Response Compensation and Liability Act of 1980 and the Hanford Federal Facility Agreement and Consent Order. Due to progressive physical deterioration of this facility, a decontamination and decommissioning plan is being considered for the immediate future. This safety analysis describes the proposed actions involved in this D ampersand D effort; identifies the radioactive material inventories involved; reviews site specific environmental characteristics and postulates an accident scenario that is evaluated to identify resultant effects

  18. Special characteristics of the safety analysis of BWRs

    International Nuclear Information System (INIS)

    The boiling water reactor uses the direct cycle. The live steam with 70 bar saturated pressure is given directly to the turbine. The reactor coolant recirculation is performed by means of internal pumps which are directly attached to the vessel bottom. This general arrangement offers some advantages concerning safety analysis. For example, in the case of a loss of coolant accident, the reactor pressure vessel itself forms a refloodable tank, which results, by proper design of the emergency cooling systems, in only a small increase of fuel temperature. (orig./RW)

  19. An Operational Safety and Health Program.

    Science.gov (United States)

    Uhorchak, Robert E.

    1983-01-01

    Describes safety/health program activities at Research Triangle Institute (North Carolina). These include: radioisotope/radiation and hazardous chemical/carcinogen use, training, monitoring, disposal; chemical waste management; air monitoring and analysis; medical program; fire safety/training, including emergency planning; Occupational Safety and…

  20. Chemical Diversity, Origin, and Analysis of Phycotoxins.

    Science.gov (United States)

    Rasmussen, Silas Anselm; Andersen, Aaron John Christian; Andersen, Nikolaj Gedsted; Nielsen, Kristian Fog; Hansen, Per Juel; Larsen, Thomas Ostenfeld

    2016-03-25

    Microalgae, particularly those from the lineage Dinoflagellata, are very well-known for their ability to produce phycotoxins that may accumulate in the marine food chain and eventually cause poisoning in humans. This includes toxins accumulating in shellfish, such as saxitoxin, okadaic acid, yessotoxins, azaspiracids, brevetoxins, and pinnatoxins. Other toxins, such as ciguatoxins and maitotoxins, accumulate in fish, where, as is the case for the latter compounds, they can be metabolized to even more toxic metabolites. On the other hand, much less is known about the chemical nature of compounds that are toxic to fish, the so-called ichthyotoxins. Despite numerous reports of algal blooms causing massive fish kills worldwide, only a few types of compounds, such as the karlotoxins, have been proven to be true ichthyotoxins. This review will highlight marine microalgae as the source of some of the most complex natural compounds known to mankind, with chemical structures that show no resemblance to what has been characterized from plants, fungi, or bacteria. In addition, it will summarize algal species known to be related to fish-killing blooms, but from which ichthyotoxins are yet to be characterized. PMID:26901085