WorldWideScience

Sample records for plant safety analysis

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zurron-Cifuentes, Oscar; Ortiz-Trujillo, Diego; Blanco-Fernandez, Luis A. [ENUSA Industrias Avanzadas S. A., Juzbado Nuclear Fuel Fabrication Plant, Ctra. Salamanca-Ledesma, km. 26, 37015 Juzbado, Salamanca (Spain)

    2010-07-01

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

  3. 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.

  4. Technology development of maintenance optimization and reliability analysis for safety features in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Choi, Seong Soo; Lee, Dong Gue; Kim, Young Il

    1999-12-01

    The reliability data management system (RDMS) for safety systems of PHWR type plants has been developed and utilized in the reliability analysis of the special safety systems of Wolsong Unit 1,2 with plant overhaul period lengthened. The RDMS is developed for the periodic efficient reliability analysis of the safety systems of Wolsong Unit 1,2. In addition, this system provides the function of analyzing the effects on safety system unavailability if the test period of a test procedure changes as well as the function of optimizing the test periods of safety-related test procedures. The RDMS can be utilized in handling the requests of the regulatory institute actively with regard to the reliability validation of safety systems. (author)

  5. Analysis and insights from a dynamical model of nuclear plant safety risk

    Energy Technology Data Exchange (ETDEWEB)

    Hess, Stephen M. [Electric Power Research Institute, 30 Bethel Road, Glen Mills, PA 19342 (United States)]. E-mail: shess@epri.com; Albano, Alfonso M. [School of Economics and Social Sciences, Singapore Management University, 90 Stamford Road, Singapore 178903 (Singapore); Gaertner, John P. [Electric Power Research Institute, 1300 Harris Boulevard, Charlotte, NC 28262 (United States)

    2007-01-15

    In this paper, we expand upon previously reported results of a dynamical systems model for the impact of plant processes and programmatic performance on nuclear plant safety risk. We utilize both analytical techniques and numerical simulations typical of the analysis of nonlinear dynamical systems to obtain insights important for effective risk management. This includes use of bifurcation diagrams to show that period doubling bifurcations and regions of chaotic dynamics can occur. We also investigate the impact of risk mitigating functions (equipment reliability and loss prevention) on plant safety risk and demonstrate that these functions are capable of improving risk to levels that are better than those that are represented in a traditional risk assessment. Next, we analyze the system response to the presence of external noise and obtain some conclusions with respect to the allocation of resources to ensure that safety is maintained at optimal levels. In particular, we demonstrate that the model supports the importance of management and regulator attention to plants that have demonstrated poor performance by providing an external stimulus to obtain desired improvements. Equally important, the model suggests that excessive intervention, by either plant management or regulatory authorities, can have a deleterious impact on safety for plants that are operating with very effective programs and processes. Finally, we propose a modification to the model that accounts for the impact of plant risk culture on process performance and plant safety risk. We then use numerical simulations to demonstrate the important safety benefits of a strong risk culture.

  6. An integrated safety for business analysis of process plants

    NARCIS (Netherlands)

    Reinders, J.E.A.; Gort, J.; Kamperveen, J.P.; Zwanikken, S.L.J.

    2004-01-01

    In this study a safety assessment technique is described, in which the contributions of a number of technical and organisational aspects of safety can be compared and quantified. Results can be used for input in recommendations regarding investment decisions, and for development of safety performanc

  7. B Plant interim safety basis

    Energy Technology Data Exchange (ETDEWEB)

    Chalk, S.E.

    1996-09-01

    This interim safety basis (ISB-008) replaces the B Plant Safety Analysis Report, WHC-SD-WM-SAR-013, Rev. 2 (WHC 1993a). ISB-008 uses existing accident analyses, modified existing accident analyses, and new accident analyses to prove that B Plant remains within the safety envelope for transition, deactivation, standby, and shutdown activities. The analyses in ISB-008 are in accordance with the most current requirements for analytical approach, risk determination, and configuration management. This document and supporting accident analyses replace previous design-basis documents.

  8. A probabilistic safety analysis of UF{sub 6} handling at the Portsmouth Gaseous Diffusion Plant

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, G.J.; Lewis, S.R.; Summitt, R.L. [Safety and Reliability Optimization Services (SAROS), Inc., Knoxville, TN (United States)

    1991-12-31

    A probabilistic safety study of UF{sub 6} handling activities at the Portsmouth Gaseous Diffusion Plant has recently been completed. The analysis provides a unique perspective on the safety of UF{sub 6} handling activities. The estimated release frequencies provide an understanding of current risks, and the examination of individual contributors yields a ranking of important plant features and operations. Aside from the probabilistic results, however, there is an even more important benefit derived from a systematic modeling of all operations. The integrated approach employed in the analysis allows the interrelationships among the equipment and the required operations to be explored in depth. This paper summarizes the methods used in the study and provides an overview of some of the technical insights that were obtained. Specific areas of possible improvement in operations are described.

  9. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  10. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  11. Analysis of safety information for nuclear power plants and development of source term estimation program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Choi, Seong Soo; Park, Jin Hee [Atomic Creative Technology, Taejon (Korea, Republic of)

    1999-12-15

    Current CARE(Computerized Advisory System for Radiological Emergency) in KINS(Korea Institute of Nuclear Safety) has no STES(Source Term Estimation System) which links between SIDS(Safety Information Display System) and FADAS(Following Accident Dose Assessment System). So in this study, STES is under development. STES system is the system that estimates the source term based on the safety information provided by SIDS. Estimated source term is given to FADAS as an input for estimation of environmental effect of radiation. Through this first year project STES for the Kori 3,4 and Younggwang 1,2 has been developed. Since there is no CARE for Wolsong(PHWR) plants yet, CARE for Wolsong is under construction. The safety parameters are selected and the safety information display screens and the alarm logic for plant status change are developed for Wolsong Unit 2 based on the design documents for CANDU plants.

  12. Food Safety and Plant Health in Ghana - Analysis of the Sanitary and Phytosanitary

    NARCIS (Netherlands)

    Maden, van der E.C.L.J.; Tay, J.G.; Koomen, I.

    2014-01-01

    Ghana is a signatory of the WTO agreement on Sanitary and Phytosanitary (SPS) measures. This international agreement sets out a framework for food safety as well as plant health. While food safety is important for both national as well as international trade, phytosanitary compliance can especially

  13. Process and plant safety

    CERN Document Server

    Hauptmanns, Ulrich

    2015-01-01

    Accidents in technical installations are random events. Hence they cannot be totally avoided. Only the probability of their occurrence may be reduced and their consequences be mitigated. The book proceeds from hazards caused by materials and process conditions to indicating technical and organizational measures for achieving the objectives of reduction and mitigation. Qualitative methods for identifying weaknesses of design and increasing safety as well as models for assessing accident consequences are presented. The quantitative assessment of the effectiveness of safety measures is explained. The treatment of uncertainties plays a role there. They stem from the random character of the accident and from lacks of knowledge on some of the phenomena to be addressed. The reader is acquainted with the simulation of accidents, safety and risk analyses and learns how to judge the potential and limitations of mathematical modelling. Risk analysis is applied amongst others to “functional safety” and the determinat...

  14. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report. Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    Herborn, D.I.

    1993-11-01

    Westinghouse Hanford Company (WHC) is the Integrating Contractor for the Hanford Waste Vitrification Plant (HWVP) Project, and as such is responsible for preparation of the HWVP Preliminary Safety Analysis Report (PSAR). The HWVP PSAR was prepared pursuant to the requirements for safety analyses contained in US Department of Energy (DOE) Orders 4700.1, Project Management System (DOE 1987); 5480.5, Safety of Nuclear Facilities (DOE 1986a); 5481.lB, Safety Analysis and Review System (DOE 1986b) which was superseded by DOE order 5480-23, Nuclear Safety Analysis Reports, for nuclear facilities effective April 30, 1992 (DOE 1992); and 6430.lA, General Design Criteria (DOE 1989). The WHC procedures that, in large part, implement these DOE requirements are contained in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals.

  15. Pantex Plant final safety analysis report, Zone 4 magazines. Staging or interim storage for nuclear weapons and components: Issue D

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    This Safety Analysis Report (SAR) contains a detailed description and evaluation of the significant environmental, safety, and health (ES&H) issues associated with the operations of the Pantex Plant modified-Richmond and steel arch construction (SAC) magazines in Zone 4. It provides (1) an overall description of the magazines, the Pantex Plant, and its surroundings; (2) a systematic evaluations of the hazards that could occur as a result of the operations performed in these magazines; (3) descriptions and analyses of the adequacy of the measures taken to eliminate, control, or mitigate the identified hazards; and (4) analyses of potential accidents and their associated risks.

  16. Safety analysis -- 200 Area Savannah River Plant, F-Canyon Operations. Supplement 4

    Energy Technology Data Exchange (ETDEWEB)

    Beary, M.M.; Collier, C.D.; Fairobent, L.A.; Graham, R.F.; Mason, C.L.; McDuffee, W.T.; Owen, T.L.; Walker, D.H.

    1986-02-01

    The F-Canyon facility is located in the 200 Separations Area and uses the Purex process to recover plutonium from reactor-irradiated uranium. The irradiated uranium is normally in the form of solid or hollow cylinders called slugs. These slugs are encased in aluminum cladding and are sent to the F-Canyon from the Savannah River Plant (SRP) reactor areas or from the Receiving Basin for Offsite Fuels (RBOF). This Safety Analysis Report (SAR) documents an analysis of the F-Canyon operations and is an update to a section of a previous SAR. The previous SAR documented an analysis of the entire 200 Separations Area operations. This SAR documents an analysis of the F-Canyon and is one of a series of documents for the Separations Area as specified in the Savannah River Implementation Plans. A substantial amount of the information supporting the conclusions of this SAR is found in the Systems Analysis. Some F-Canyon equipment has been updated during the time between the Systems Analysis and this SAR and a complete description of this equipment is included in this report. The primary purpose of the analysis was to demonstrate that the F-Canyon can be operated without undue risk to onsite or offsite populations and to the environment. In this report, risk is defined as the expected frequency of an accident, multiplied by the resulting radiological consequence in person-rem. The units of risk for radiological dose are person-rem/year. Maximum individual exposure values have also been calculated and reported.

  17. A survey on reliability and safety analysis techniques of robot systems in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Eom, H.S.; Kim, J.H.; Lee, J.C.; Choi, Y.R.; Moon, S.S

    2000-12-01

    The reliability and safety analysis techniques was surveyed for the purpose of overall quality improvement of reactor inspection system which is under development in our current project. The contents of this report are : 1. Reliability and safety analysis techniques suvey - Reviewed reliability and safety analysis techniques are generally accepted techniques in many industries including nuclear industry. And we selected a few techniques which are suitable for our robot system. They are falut tree analysis, failure mode and effect analysis, reliability block diagram, markov model, combinational method, and simulation method. 2. Survey on the characteristics of robot systems which are distinguished from other systems and which are important to the analysis. 3. Survey on the nuclear environmental factors which affect the reliability and safety analysis of robot system 4. Collection of the case studies of robot reliability and safety analysis which are performed in foreign countries. The analysis results of this survey will be applied to the improvement of reliability and safety of our robot system and also will be used for the formal qualification and certification of our reactor inspection system.

  18. Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

    Science.gov (United States)

    Yi, Ping; Wang, Qingkang; Kong, Xianjing

    2017-01-01

    The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

  19. 78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

    Science.gov (United States)

    2013-01-22

    ... subsection to NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power... incorporate the final approved guidance into the next revision of NUREG 0800. Dated at Rockville,...

  20. Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

    Energy Technology Data Exchange (ETDEWEB)

    ULLAH, M K

    2001-02-26

    The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

  1. Structural Safety Analysis Based on Seismic Service Conditions for Butterfly Valves in a Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Sang-Uk Han

    2014-01-01

    Full Text Available The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements.

  2. Structural safety analysis based on seismic service conditions for butterfly valves in a nuclear power plant.

    Science.gov (United States)

    Han, Sang-Uk; Ahn, Dae-Gyun; Lee, Myeong-Gon; Lee, Kwon-Hee; Han, Seung-Ho

    2014-01-01

    The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements.

  3. Operational safety analysis of the Olkiluoto encapsulation plant and disposal facility; Olkiluodon kapselointi- ja loppusijoituslaitoksen kaeyttoeturvallisuusanalyysi

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, J.; Suolanen, V. [VTT Technical Research Centre of Finland, Espoo (Finland)

    2012-11-15

    Radiation doses for workers of the facility, for inhabitants in the environment and for terrestrial ecosystem possibly caused by the encapsulation and disposal facilities to be built at Olkiluoto during its operation were considered in the study. The study covers both the normal operation of the plant and some hypothetical incidents and accidents. Release through the ventilation stack is assumed to be filtered both in normal operation and in hypothetical abnormal fault and accident cases. In addition the results for unfiltered releases are also presented. This research is limited to the deterministic analysis. During about 30 operation years of our four nuclear power plant units there have been found 58 broken fuel pins. Roughly estimating there has been one fuel leakage per year in a facility (includes two units). Based on this and adopting a conservative approach, it is estimated that one fuel pin per year could leak in normal operation during encapsulation process. The release magnitude in incidents and accidents is based on the event chains, which lead to loss of fuel pin tightness followed by a discharge of radionuclides into the handling space and to some degree to the atmosphere through the ventilation stack equipped with redundant filters. The most exposed group of inhabitants is conservatively assumed to live at the distance of 200 meters from the encapsulation and disposal plant and it will receive the largest doses in most dispersion conditions. The dose value to a member of the most exposed group was calculated on the basis of the weather data in such a way that greater dose than obtained here is caused only in 0.5 percent of dispersion conditions. The results obtained indicate that during normal operation the doses to workers remain small and the dose to the member of the most exposed group is less than 0.001 mSv per year. In the case of hypothetical fault and accident releases the offsite doses do not exceed either the limit values set by the safety

  4. Human Factors engineering criteria and design for the Hanford Waste Vitrification Plant preliminary safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Wise, J.A.; Schur, A.; Stitzel, J.C.L.

    1993-09-01

    This report provides a rationale and systematic methodology for bringing Human Factors into the safety design and operations of the Hanford Waste Vitrification Plant (HWVP). Human Factors focuses on how people perform work with tools and machine systems in designed settings. When the design of machine systems and settings take into account the capabilities and limitations of the individuals who use them, human performance can be enhanced while protecting against susceptibility to human error. The inclusion of Human Factors in the safety design of the HWVP is an essential ingredient to safe operation of the facility. The HWVP is a new construction, nonreactor nuclear facility designed to process radioactive wastes held in underground storage tanks into glass logs for permanent disposal. Its design and mission offer new opposites for implementing Human Factors while requiring some means for ensuring that the Human Factors assessments are sound, comprehensive, and appropriately directed.

  5. Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Hao-Tzu [Institute of Nuclear Energy Research, Taoyuan, Taiwan (China). Research Atomic Energy Council; Li, Wan-Yun; Wang, Jong-Rong; Tseng, Yung-Shin; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

    2017-03-15

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

  6. Robots and plant safety

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, P.

    1996-02-01

    The application of robots in the harsh environments in which TELEMAN equipment will have to operate has large benefits, but also some drawbacks. The main benefit is the ability gained to perform tasks where people cannot go, while there is a possibility of inflicting damage to the equipment handled by the robot, and the plant when mobile robots are involved. The paper describes the types of possible damage and the precautions to be taken in order to reduce the frequency of the damaging events. A literature study for the topic only gave some insight into examples, but no means for a systematic treatment of the topic. (au) 16 refs.

  7. Suitability review of FMEA and reliability analysis for digital plant protection system and digital engineered safety features actuation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, I. S.; Kim, T. K.; Kim, M. C.; Kim, B. S.; Hwang, S. W.; Ryu, K. C. [Hanyang Univ., Seoul (Korea, Republic of)

    2000-11-15

    Of the many items that should be checked out during a review stage of the licensing application for the I and C system of Ulchin 5 and 6 units, this report relates to a suitability review of the reliability analysis of Digital Plant Protection System (DPPS) and Digital Engineered Safety Features Actuation System (DESFAS). In the reliability analysis performed by the system designer, ABB-CE, fault tree analysis was used as the main methods along with Failure Modes and Effect Analysis (FMEA). However, the present regulatory technique dose not allow the system reliability analysis and its results to be appropriately evaluated. Hence, this study was carried out focusing on the following four items ; development of general review items by which to check the validity of a reliability analysis, and the subsequent review of suitability of the reliability analysis for Ulchin 5 and 6 DPPS and DESFAS L development of detailed review items by which to check the validity of an FMEA, and the subsequent review of suitability of the FMEA for Ulchin 5 and 6 DPPS and DESFAS ; development of detailed review items by which to check the validity of a fault tree analysis, and the subsequent review of suitability of the fault tree for Ulchin 5 and 6 DPPS and DESFAS ; an integrated review of the safety and reliability of the Ulchin 5 and 6 DPPS and DESFAS based on the results of the various reviews above and also of a reliability comparison between the digital systems and the comparable analog systems, i.e., and analog Plant Protection System (PPS) and and analog Engineered Safety Features Actuation System (ESFAS). According to the review mentioned above, the reliability analysis of Ulchin 5 and 6 DPPS and DESFAS generally satisfies the review requirements. However, some shortcomings of the analysis were identified in our review such that the assumed test periods for several equipment were not properly incorporated in the analysis, and failures of some equipment were not included in the

  8. Preliminary safety analysis of a PBMR supplying process heat to a co-located ethylene production plant

    Energy Technology Data Exchange (ETDEWEB)

    Scarlat, Raluca O., E-mail: rscarlat@nuc.berkeley.edu [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States); Cisneros, Anselmo T. [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States); Koutchesfahani, Tawni [University of California, Chemical and Biomolecular Engineering, 201 Gilman Hall, Berkeley, CA 94720 (United States); Hong, Rada; Peterson, Per F. [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States)

    2012-10-15

    This paper considers the safety analysis and licensing approach for co-locating a pebble bed modular reactor (PBMR) to provide process heat to an ethylene production unit. The PBMR is an advanced nuclear reactor design that provides 400 MW of thermal energy. Ethylene production is an energy intensive process that utilizes large gas furnaces to provide the heat for the process. Coupling a PBMR with an ethylene production plant would open a new market for nuclear power, and would provide the chemical industry with a cleaner power source, helping to achieve the Clean Air Act standards, and eliminating the 0.5 ton of CO{sub 2} emissions per ton of produced ethylene. Our analysis uses the Chevron Phillips chemical plant in Sweeney, TX as a prototypical site. The plant has four ethylene production trains, with a total power consumption of 2.4 GW, for an ethylene output of 3.7 million tons per year, 4% of the global ethylene production capacity. This paper proposes replacement of the gas furnaces by low-emission PBMR modules, and presents the safety concerns and risk mitigation and management options for this coupled system. Two coupling design options are proposed, and the necessary changes to the design basis events and severe accidents for the PBMR licensing application are discussed. A joint effort between the chemical and the nuclear entities to optimize the coupling design, establish preventive maintenance procedures, and develop emergency response plans for both of the units is recommended.

  9. Plutonium finishing plant safety systems and equipment list

    Energy Technology Data Exchange (ETDEWEB)

    Bergquist, G.G.

    1995-01-06

    The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex.

  10. Software safety hazard analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1996-02-01

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

  11. Safety analysis, 200 Area, Savannah River Plant H-Canyon operations. Supplement 5

    Energy Technology Data Exchange (ETDEWEB)

    Beary, M M; Collier, C D; Fairobent, L A; Graham, R F; Mason, C L; McDuffee, W T; Owen, T L; Walker, D H [Science Applications International Corp., San Diego, CA (United States)

    1986-02-01

    The H-Canyon facility is located in the 200 Separations Area and uses the HM process to separate uranium, neptunium, plutonium, and fission products. Irradiated uranium fuels containing {sup 235}U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium isotopes. This Safety Analysis Report (SAR) documents an analysis of the H-Canyon operations and is an update to a section of a previous SAR. This SAR documents an analysis of the H-Canyon and is one of a series of documents for the Separations Area as specified in the Savannah River Implementation Plans. A substantial amount of the information supporting the Conclusions of this SAR is found in the Systems Analysis. Some H-Canyon equipment has been updated during the time between the Systems Analysis and this SAR and a complete description of this equipment is included in this report. The primary purpose of the analysis was to demonstrate that the H-Carbon can be operated without due risk to onsite or offsite populations and to the environment. In this report, risk is defined an the expected frequency of an accident, multiplied by the resulting radiological consequence in person-rem. The units of risk for radiological does are person-rem/year. Maximum individual exposure values have also been calculated and reported.

  12. Probabilistic analysis of safety in industrial irradiation plants; Analisis probabilistico de seguridad en plantas industriales de irradiacion

    Energy Technology Data Exchange (ETDEWEB)

    Alderete, F.; Elechosa, C. [Autoridad Regulatoria Nuclear, Av. del Libertador 8250 - Buenos Aires (Argentina)]. e-mail: falderet@sede.arn.gov.ar

    2006-07-01

    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{sup -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

  13. Plutonium Finishing Plant safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE`s independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91.

  14. Standard model for the safety analysis report of nuclear fuel reprocessing plants; Modelo padrao para relatorio de analise de seguranca de usinas de reprocessamento de combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-02-15

    This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization.

  15. Conceptual design of an integrated information system for safety related analysis of nuclear power plants (IRIS Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Hofer, K.; Zehnder, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Galperin, A. [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

    1994-01-01

    This report deals with a conceptual design of an integrated information management system, called PSI-IRIS, as needed to assist the analysts for reactor safety related investigations on Swiss nuclear power plants within the project STARS. Performing complicated engineering analyses of an NPP requires storage and manipulation of a large amount of information, both data and knowledge. This information is characterized by its multi-disciplinary nature, complexity, and diversity. The problems caused by inefficient and lengthy manual operations involving the data flow management within the framework of the safety related analysis of an NPP, can be solved by applying computer aided engineering (CAE) principles. These principles are the basis for the design of the integrated information management system PSI-IRIS presented in this report. The basic idea is to create a computerized environment, which includes both database and functional capabilities. The database of the PSI-IRIS consists of two parts, an NPP generic database (GDB) and a collection of analysis results (CASE{sub L}IB). The GDB includes all technical plant data and information needed to generate input decks for all computer codes utilized within the STARS project. The CASE{sub L}IB storage contains the accumulated knowledge, input decks, and result files of the NPP transient analyses. Considerations and analysis of the data types and the required data manipulation capabilities as well as operational requirements resulted in the choice of an object-oriented database management system (OODBMS) as a development platform for solving the software engineering problems. Several advantages of OODBMS`s over conventional relational database management systems were found of crucial importance, especially providing the necessary flexibility for different data types and the potential for extensibility. (author) 15 figs., tabs., 20 refs.

  16. Preliminary systems-interaction results from the Digraph Matrix Analysis of the Watts Bar Nuclear Power Plant safety-injection systems

    Energy Technology Data Exchange (ETDEWEB)

    Sacks, I.J.; Ashmore, B.C.; Champney, J.M.; Alesso, H.P.

    1983-06-01

    This report provides preliminary results generated by a Digraph Matrix Analysis (DMA) for a Systems Interaction analysis performed on the Safety Injection System of the Tennessee Valley Authority Watts Bar Nuclear Power Plant. An overview of DMA is provided along with a brief description of the computer codes used in DMA.

  17. Preliminary systems-interaction results from the Digraph Matrix Analysis of the Watts Bar Nuclear Power Plant safety-injection systems

    Energy Technology Data Exchange (ETDEWEB)

    Sacks, I.J.; Ashmore, B.C.; Champney, J.M.; Alesso, H.P.

    1983-06-01

    This report provides preliminary results generated by a Digraph Matrix Analysis (DMA) for a Systems Interaction analysis performed on the Safety Injection System of the Tennessee Valley Authority Watts Bar Nuclear Power Plant. An overview of DMA is provided along with a brief description of the computer codes used in DMA.

  18. EDITORIAL: Safety aspects of fusion power plants

    Science.gov (United States)

    Kolbasov, B. N.

    2007-07-01

    importance for the fusion power plant research programmes. The objective of this Technical Meeting was to examine in an integrated way all the safety aspects anticipated to be relevant to the first fusion power plant prototype expected to become operational by the middle of the century, leading to the first generation of economically viable fusion power plants with attractive S&E features. After screening by guest editors and consideration by referees, 13 (out of 28) papers were accepted for publication. They are devoted to the following safety topics: power plant safety; fusion specific operational safety approaches; test blanket modules; accident analysis; tritium safety and inventories; decommissioning and waste. The paper `Main safety issues at the transition from ITER to fusion power plants' by W. Gulden et al (EU) highlights the differences between ITER and future fusion power plants with magnetic confinement (off-site dose acceptance criteria, consequences of accidents inside and outside the design basis, occupational radiation exposure, and waste management, including recycling and/or final disposal in repositories) on the basis of the most recent European fusion power plant conceptual study. Ongoing S&E studies within the US inertial fusion energy (IFE) community are focusing on two design concepts. These are the high average power laser (HAPL) programme for development of a dry-wall, laser-driven IFE power plant, and the Z-pinch IFE programme for the production of an economically-attractive power plant using high-yield Z-pinch-driven targets. The main safety issues related to these programmes are reviewed in the paper `Status of IFE safety and environmental activities in the US' by S. Reyes et al (USA). The authors propose future directions of research in the IFE S&E area. In the paper `Recent accomplishments and future directions in the US Fusion Safety & Environmental Program' D. Petti et al (USA) state that the US fusion programme has long recognized that the S

  19. 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.

  20. Safety analysis, 200 Area, Savannah River Plant: Separations area operations. Receiving Basin for Offsite Fuel (Supplement 3)

    Energy Technology Data Exchange (ETDEWEB)

    Allen, P M

    1983-09-01

    Analysis of the Savannah River Plant RBOF and RRF included an evaluation of the reliability of process equipment and controls, administrative controls, and engineered safety features. The evaluation also identified potential scenarios and radiological consequences. Risks were calculated in terms of 50-year population dose commitment per year (man-rem/year) to the onsite and offsite population within an 80 Km radius of RBOF and RRF, and to an individual at the plant boundary. The total 50-year onsite and offsite population radiological risks of operating the RBOF and RRF were estimated to be 1.0 man-rem/year. These risks are significantly less than the population dose of 54,000 man/rem/yr for natural background radiation in a 50-mile radius. The 50-year maximum offsite individual risk from operating the facility was estimated to be 2.1 {times} 10{sup 5} rem/yr. These risks are significantly lower than 93 mrem/yr an individual is expected to receive from natural background radiation in this area. The analysis shows. that the RBOF and RRF can be operated without undue risk to onsite personnel or to the general public.

  1. Probabilistic safety assessment for optimum nuclear power plant life management (PLiM) theory and application of reliability analysis methods for major power plant components

    CERN Document Server

    Arkadov, G V; Rodionov, A N

    2012-01-01

    Probabilistic safety assessment methods are used to calculate nuclear power plant durability and resource lifetime. Directing preventative maintenance, this title provides a comprehensive review of the theory and application of these methods.$bProbabilistic safety assessment methods are used to calculate nuclear power plant durability and resource lifetime. Successful calculation of the reliability and ageing of components is critical for forecasting safety and directing preventative maintenance, and Probabilistic safety assessment for optimum nuclear power plant life management provides a comprehensive review of the theory and application of these methods. Part one reviews probabilistic methods for predicting the reliability of equipment. Following an introduction to key terminology, concepts and definitions, formal-statistical and various physico-statistical approaches are discussed. Approaches based on the use of defect-free models are considered, along with those using binomial distribution and models bas...

  2. Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-18

    This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met.

  3. Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Chang, T.Y.

    1991-09-01

    Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.

  4. Final Safety Analysis Addenda to Hazards Summary Report, Experimental Breeder Reactor II (EBR-II): upgrading of plant protection system. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    Allen, N. L.; Keeton, J. M.; Sackett, J. I. [comps.

    1980-06-01

    This report is the second in a series of compilations of the formal Final Safety Analysis Addenda (FSAA`s) to the EBR-II Hazard Summary Report and Addendum. Sections 2 and 3 are edited versions of the original FSAA`s prepared in support of certain modifications to the reactor-shutdown-system portion of the EBR-II plant-protection system. Section 4 is an edited version of the original FSAA prepared in support of certain modifications to a system classified as an engineered safety feature. These sections describe the pre- and postmodification system, the rationale for the modification, and required supporting safety analysis. Section 5 provides an updated description and analysis of the EBR-II emergency power system. Section 6 summarizes all significant modifications to the EBR-II plant-protection system to date.

  5. Systems Thinking Safety Analysis: Nuclear Security Assessment of Physical Protection System in Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Tae Ho Woo

    2013-01-01

    Full Text Available The dynamical assessment has been performed in the aspect of the nuclear power plants (NPPs security. The physical protection system (PPS is constructed by the cyber security evaluation tool (CSET for the nuclear security assessment. The systems thinking algorithm is used for the quantifications by the Vensim software package. There is a period of 60 years which is the life time of NPPs' operation. The maximum possibility happens as 3.59 in the 30th year. The minimum value is done as 1.26 in the 55th year. The difference is about 2.85 times. The results of the case with time delay have shown that the maximum possibility of terror or sabotage incident happens as 447.42 in the 58th year and the minimum value happens as 89.77 in the 51st year. The difference is about 4.98 times. Hence, if the sabotage happens, the worst case is that the intruder can attack the target of the nuclear material in about one and a half hours. The general NPPs are modeled in the study and controlled by the systematic procedures.

  6. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

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

  7. Component- and plant safety. Bauteil- und Anlagensicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R. (Fachhochschule Koeln (Germany). Fachbereich Anlagen- und Verfahrenstechnik)

    1992-01-01

    Since reliable plants depend on reliable components, the work of engineers entrusted with their planning or maintenance would be unthinkable today without a sufficiently founded safety-engineering knowledge. This book gives a profound introduction to component and plant safety and discusses important licensing aspects. It mainly addresses students of mechanical engineering, chemical engineering, supply and environmental engineering and experienced industrial-planning experts who have some elementary background knowledge of mathematics, mechanics, material science and plant construction. (orig.).

  8. Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches

    Energy Technology Data Exchange (ETDEWEB)

    Steven R. Sherman

    2007-06-01

    The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

  9. 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.)

  10. Safety in nuclear power plants in India

    OpenAIRE

    Deolalikar R

    2008-01-01

    Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operat...

  11. A study on safety assessment methodology for a vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

    2002-03-15

    In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

  12. Mathematical Safety Assessment Approaches for Thermal Power Plants

    Directory of Open Access Journals (Sweden)

    Zong-Xiao Yang

    2014-01-01

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

  13. Probabilistic analysis of safety of a production plant of hydrogen using nuclear energy; Analisis probabilistico de seguridad de una planta de produccion de hidrogeno utilizando energia nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Flores F, A. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico); Nelson E, P.F.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: alain_fyf@yahoo.com

    2005-07-01

    The present work makes use of the Probabilistic Safety analysis to evaluate and to quantify the safety in a plant producer of hydrogen coupled to a nuclear reactor of high temperature, the one which is building in Japan. It is had the description of systems and devices of the HTTR, the pipe diagrams and instrumentation of the plant, as well as the rates of generic faults for the components of the plant. The first step was to carry out a HAZOP study (Hazard and Operability Study) with the purpose of obtaining the initiator events; once obtained these, it was developed a tree of events by each initiator event and for each system it was developed a fault tree; the data used for the quantification of the failure probability of the systems were obtained starting from several generic sources of information. In each tree of events different final states were obtained and it stops each one, their occurrence frequency. The construction and evaluation of the tree of events and of failures one carries out with the SAPHIRE program. The results show the safety of the shutdown system of the HTTR and they allow to suggest modifications to the auxiliary system of refrigeration and to the heat exchanger helium/water pressurized. (Author)

  14. Systems interaction results from the digraph matrix analysis of the Watts Bar Nuclear Power Plant high pressure safety injection systems. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Sacks, I.J.; Ashmore, B.C.; Alesso, H.P.

    1983-07-01

    Spatial and functional coupling (including human actions) of nuclear power plant systems that lead to interdependencies are called Systems Interactions. At present, the US Nuclear Regulatory Commission (NRC) is investigating ways of integrating a systems interactions study with existing Probabilistic Risk Assessment efforts. One approach is based on graph-theoretic methods utilizing matrix representations of logic diagrams called Digraph Matrix Analysis (DMA). The objective in this report is to demonstrate the capabilities of Digraph Matrix Analysis to model an accident sequence (including front-line systems, support systems and human actions) as a continuous, well-integrated logic model in order to identify and evaluate functional systems interactions. The selected accident sequence, loss of high pressure safety injection during a LOCA, was modeled and qualitative and quantitative comparisons were made to the Reactor Safety Study (WASH 1400) and other studies.

  15. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  16. Safety Assessment - Swedish Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B. [Luleaa Univ. of Technology (Sweden)

    1996-12-31

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs.

  17. Integrated Safety Analysis Tiers

    Science.gov (United States)

    Shackelford, Carla; McNairy, Lisa; Wetherholt, Jon

    2009-01-01

    Commercial partnerships and organizational constraints, combined with complex systems, may lead to division of hazard analysis across organizations. This division could cause important hazards to be overlooked, causes to be missed, controls for a hazard to be incomplete, or verifications to be inefficient. Each organization s team must understand at least one level beyond the interface sufficiently enough to comprehend integrated hazards. This paper will discuss various ways to properly divide analysis among organizations. The Ares I launch vehicle integrated safety analyses effort will be utilized to illustrate an approach that addresses the key issues and concerns arising from multiple analysis responsibilities.

  18. Safety assessment of plant food supplements (PFS)

    NARCIS (Netherlands)

    Berg, van den S.J.P.L.; Serra-Majem, L.; Coppens, P.; Rietjens, I.

    2011-01-01

    Botanicals and botanical preparations, including plant food supplements (PFS), are widely used in Western diets. The growing use of PFS is accompanied by an increasing concern because the safety of these PFS is not generally assessed before they enter the market. Regulatory bodies have become more a

  19. Safety in nuclear power plants in India

    Directory of Open Access Journals (Sweden)

    Deolalikar R

    2008-01-01

    Full Text Available Safety in nuclear power plants (NPPs in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements.

  20. Safety in nuclear power plants in India.

    Science.gov (United States)

    Deolalikar, R

    2008-12-01

    Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements.

  1. Development of specific data of plant for a safety probabilistic analysis; Desarrollo de datos especificos de planta para un analisis probabilistico de seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, M. [Emersis S.A. de C.V., Tabachines 9-bis, 62589 Temixco, Morelos (Mexico); Nelson E, P. [LAIRN, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: cuesta@emersis.com

    2004-07-01

    In this work the development of specific data of plant is described for the Safety Probabilistic Analysis (APS) of the Laguna Verde Central. The description of those used methods concentrate on the obtention of rates of failure of the equipment and frequencies of initiator events modeled in the APS, making mention to other types of data that also appeal to specific sources of the plant. The method to obtain the rates of failure of the equipment takes advantage the information of failures of components and unavailability of systems obtained entreaty in execution with the Maintenance Rule (1OCFR50.65). The method to develop the frequencies of initiators take in account the registered operational experience as reportable events. In both cases the own experience is combined with published generic data using Bayesian realized techniques. Details are provided about the gathering of information, the confirmations of consistency and adjustment necessities, presenting examples of the obtained results. (Author)

  2. CONVEYOR SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

    The purpose and objective of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) surface and subsurface conveyor system (for a list of conveyor subsystems see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the conveyor structures/systems/components in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component (S/S/C) design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions. The scope of this analysis is limited to the hazards related to the design of conveyor structures/systems/components (S/S/Cs) that occur during normal operation. Hazards occurring during assembly, test and maintenance or ''off normal'' operations have not been included in this analysis. Construction related work activities are specifically excluded per DOE Order 5481.1B section 4. c.

  3. 合成氨厂液氨库安全因素分析及安全措施%ANALYSIS OF SAFETY FACTORS AND SAFETY MEASURES FOR TANK-FARM OF LIQUID AMMONIA IN AMMONIA PLANT

    Institute of Scientific and Technical Information of China (English)

    刘荣; 杨涛

    2012-01-01

    以某合成氨厂为例,对该合成氨厂液氨球罐可能发生的各种危险因素采用软件进行计算和分析,有针对性的在设计中提出安全设施和措施。%Both calculation and analysis for various dangerous factors possibly occurred on a tank-farm of liquid ammonia in an ammonia plant have been carded out and the targeted safety facilities measures are requested in design work..

  4. Systems interaction results from the digraph matrix analysis of the Watts Bar Nuclear Power Plant high pressure safety injection systems. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Sacks, I.J.; Ashmore, B.C.; Alesso, H.P.

    1983-07-01

    Spatial and functional coupling of nuclear power plant systems that lead to interdependencies are called Systems Interactions. At present, the US Nuclear Regulatory Commission (NRC) is investigating ways of integrating a systems interactions study with existing Probabilistic Risk Assessment efforts. One approach is based on graph-theoretic methods utilizing matrix representations of logic diagrams called Digraph Matrix Analysis (DMA). The objective in this report is to demonstrate the capabilities of Digraph Matrix Analysis to model an accident sequence (including front line systems, support systems and human actions) as a continuous, well-integrated logic model in order to identify and evaluate functional systems interactions. The selected accident sequence, loss of high pressure safety injection during an S1 LOCA, was modeled and qualitative and quantitative comparisons were made to WASH 1400 aand other studies.

  5. Deep Borehole Disposal Safety Analysis.

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  6. Safety and security aspects in design of digital safety I and C in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

    2016-05-15

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  7. Software safety analysis practice in installation phase

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  8. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

    2014-10-15

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  9. A study on the human reliability analysis in probabilistic safety assessment during low power/shutdown operation of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kang, D. I.; Sung, T. Y.; Jung, W. D.; Yang, J. E.; Park, J. H.; Lee, Y. H.; Hwang, M. J.; Kim, K. Y.; Jin, Y. H

    1997-02-01

    This report describes the review results of human reliability analysis (HRA) in the probabilistic safety assessment (PSA) during low power/shutdown operation of nuclear power plants (NPPs). We select four NPP PSA reports to review. These are System 80+ using THERP, Surry using SLIM, Grand Gulf using ASEP, Electric de France using simulator experiments. This report also describe the method, the procedure, the quantification example, the critical review and the insights of HRA which were used in the process of PSA of NPPs mentioned above. It is expected that this study results will be effectively used in HRA of domestic PSA during low power/shutdown operation of NPPs. (author). 18 refs., 1 tab.

  10. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  11. Strategic Thinking and Measures for Plant Quarantine and Food Safety

    Institute of Scientific and Technical Information of China (English)

    Guibin; WANG; Jinghan; YANG; Yongmei; LI; Daiju; CUN; Rui; MA; Guiping; FAN

    2013-01-01

    This paper describes the importance of plant quarantine to the food safety in China through cases where plant quarantine helps to effectively intercept harmful organisms for food safety and promote export and import trading.It also presents the existing problems in plant quarantine work and appropriate measures.

  12. An integrated design methodology for the safety and security of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Joung, S. Y.; Chang, S. H. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2012-03-15

    After Fukushima nuclear power plant accident, safety of nuclear power plant was issued. In Fukushima accident, one of main reason is location of emergency diesel power generators stop which locate under the sea water level when tsunami occurred. In view of security, emergency diesel generator location under reactor building design is good because for example to escape air strike but is not good for safety for example Fukushima accident. Sometimes safety and security design looks conflicting but nuclear safety and nuclear security share the goal of protecting but nuclear safety and nuclear power plants operate at acceptable risk levels. The purpose of this paper is to introduce safety and security integrated design for nuclear power plant in special emergency diesel generator and control room with simple probabilistic safety assessment analysis.

  13. Development of a dynamical systems model of plant programmatic performance on nuclear power plant safety risk

    Energy Technology Data Exchange (ETDEWEB)

    Hess, Stephen M. [Sensortex, Inc., 515 Schoolhouse Road, Kennett Square, PA 19348 (United States)]. E-mail: smhess@sensortex.com; Albano, Alfonso M. [Department of Physics, Bryn Mawr College, Bryn Mawr, PA 19010 (United States); Gaertner, John P. [Electric Power Research Institute, 1300 Harris Boulevard, Charlotte, NC 28262 (United States)

    2005-10-01

    Application of probabilistic risk assessment (PRA) techniques to model nuclear power plant accident sequences has provided a significant contribution to understanding the potential initiating events, equipment failures and operator errors that can lead to core damage accidents. Application of the lessons learned from these analyses has resulted in significant improvements in plant operation and safety. However, this approach has not been nearly as successful in addressing the impact of plant processes and management effectiveness on the risks of plant operation. The research described in this paper presents an alternative approach to addressing this issue. In this paper we propose a dynamical systems model that describes the interaction of important plant processes on nuclear safety risk. We discuss development of the mathematical model including the identification and interpretation of significant inter-process interactions. Next, we review the techniques applicable to analysis of nonlinear dynamical systems that are utilized in the characterization of the model. This is followed by a preliminary analysis of the model that demonstrates that its dynamical evolution displays features that have been observed at commercially operating plants. From this analysis, several significant insights are presented with respect to the effective control of nuclear safety risk. As an important example, analysis of the model dynamics indicates that significant benefits in effectively managing risk are obtained by integrating the plant operation and work management processes such that decisions are made utilizing a multidisciplinary and collaborative approach. We note that although the model was developed specifically to be applicable to nuclear power plants, many of the insights and conclusions obtained are likely applicable to other process industries.

  14. Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Park, Y. W.; Park, H.G. [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly.

  15. 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}.

  16. Safety analysis of DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Song, G. H.; Ahn, J. Y.; Chung, C. H. [Hyundai Engineering and Construction Ltd., Seoul (Korea)

    2000-11-01

    M6 hot cell of IMEF which is to be used for DUPIC fuel fabrication experiment was constructed as an {alpha}-{gamma} hot cell for material examination of small amount of high-burnup fuel. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and environmental evaluation results but new requirement of nuclear criticality evaluation. This study includes evaluation of shielding, environmental effect and nuclear criticality for DUPIC fuel development facility. Results of the above evaluation have verified that both national regulation limit and IMEF design criteria were satisfied. Therefore, the result of this study can be used for verification document of facility safety and, in addition, this report can be submitted as supplement document for IMEF licensing modification in accordance with DUPIC fuel fabrication. 18 refs., 10 figs., 31 tabs. (Author)

  17. Airline Safety: A Comparative Analysis.

    Science.gov (United States)

    1987-01-01

    Argentinas Olympic AUA (Austria) PIA (Pakistan) Avianca (Colombia) Pan American British Airways PAL (Philippines) East African Qantas Egyptair Sabena El...S.TP OFR O T PEIDCV E Airline Safety: A Comparative Analysis TRlES IS1j0’~fJ 6. PERFORMING 01G. REPORT NUMBER AU TNOR( ) Sign . CONTRACT OR GRANT NUMBER...OF I NOVa IS 1 OBSOLETE SECURITY CLASSIFICATION OF THIS PAGE j",.n Des. Enterod) 87 jO 1 4 Xb AIRLINE SAFETY: A COMPARATIVE ANALYSIS by Mary Katherine

  18. Forum for fire protection and safety in power plants[Norway

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The conference contains 16 presentations on topics in the fields of fire protection and safety in plants in Western Norway, reorganization and reconstruction of power systems and plants in Norway, various aspects of risk and vulnerability analysis, technological aspects of plant management and construction and problems and risks with particularly transformers. Some views on challenges of the fire departments and the new Norwegian regulations for electrical power supply systems are included. One presentation deals with challenges for Icelandic power production plants.

  19. A Study on Methodologies for Assessing Safety Critical Network's Risk Impact on Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lim, T. J.; Park, S. K.; Seo, S. J. [Soongsil Univ., Seoul (Korea, Republic of)

    2007-04-15

    The objectives of this project is to establish methodologies for assessing safety-critical network's risk impact on nuclear power plant by developing reliability analysis models for the safety-critical network. It is essential to develop reliability analysis models for safety-critical network, and it is very important to adapt the model to the current methodologies for assessing risk impact on nuclear power plants. Major outputs of the first year study are preliminary models for assessing reliability of safety-critical communication networks and those of the second year study are methodologies for assessing safety-critical network's risk impact on nuclear power plant.

  20. Questions concerning safety and risk after the nuclear accidents in Japan. Deepened accident analysis for the Fukushima Daiichi power plant; Sicherheits- und Risikofragen im Nachgang zu den nuklearen Stoer- und Unfaellen in Japan. Vertiefte Ereignisanalyse zur Anlage Fukushima-Daini

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

    2015-02-25

    The study questions concerning safety and risk in Japanese power plants following the disastrous nuclear accident covers the following issues: the nuclear facility Fukushima Daiichi, site characterization, important technical equipment, important electro-technical equipment, personal; description of the accident progression in the Fukushima nuclear power plant: impact of the earthquake, impact of the tsunami, short-term measures of the operating personnel, pressure and temperature situation in the containments, restoration of the after-heat cooling system in the units 1/2 and 4, fuel element storage pool, summarized parameters during the accident progress; comparative analysis of the accident progression at the Fukushima Daiichi site.

  1. Development and verification of a software system for the probabilistic safety analysis of nuclear plants as part of the proryv project

    Directory of Open Access Journals (Sweden)

    L.V. Abramov

    2016-06-01

    The paper presents results of the CRISS 5.3 code verification through the comparison of the analysis results obtained using the CRISS 5.3 system against analytical formulas and results of a qualitative and quantitative analysis based on certified nuclear plant PSA software tools.

  2. Improving the safety of LWR power plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs (improving or maintaining level of safety with simpler systems or in a more cost-effective manner).

  3. Safety assessment of a nuclear power plant building subjected to an aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Thai, Duc-Kien; Kim, Seung-Eock, E-mail: sekim@sejong.ac.kr

    2015-11-15

    Highlights: • Numerical analysis of a nuclear auxiliary building under aircraft crash is conducted. • The analysis result of impact force is verified using the Riera function. • The safety assessment is performed with regard to different impact scenarios. • Discussions and conclusions on safety of the nuclear building are presented. - Abstract: This paper presents a safety assessment of a nuclear building subjected to an aircraft crash using numerical analysis. For impact simulation, the reinforced concrete (RC) Primary Auxiliary Building (PAB) of the Korea Standard Nuclear Power Plant (KSNP) is fully modeled and an aircraft model of a Boeing 767-400 is used. The Riera function is used to verify the analysis result of impact force–time history. The IRIS test is used to verify the structural behavior of the RC wall under impact loading. The safety assessment of the building is performed with regard to different impact scenarios. The safety of the nuclear building under aircraft crash, including (1) global structural safety, (2) local structural safety, and (3) vibration safety are evaluated and discussed. The results show that the global and local structural safety of the PAB is ensured in all impact scenarios. However, the vibration safety of the building is not ensured. In accordance, the regulatory guide of United States Nuclear Regulatory Commission (U.S. NRC), shutdown of the nuclear power plant is required.

  4. The PEC reactor. Safety analysis: Detailed reports

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

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

  5. Correct safety requirements during the life cycle of heating plants; Korrekta saekerhetskrav under vaermeanlaeggningars livscykel

    Energy Technology Data Exchange (ETDEWEB)

    Tegehall, Jan; Hedberg, Johan [Swedish National Testing and Research Inst., Boraas (Sweden)

    2006-10-15

    The safety of old steam boilers or hot water generators is in principle based on electromechanical components which are generally easy to understand. The use of safety-PLC is a new and flexible way to design a safe system. A programmable system offers more degrees of freedom and consequently new problems may arise. As a result, new standards which use the Safety Integrity Level (SIL) concept for the level of safety have been elaborated. The goal is to define a way of working to handle requirements on safety in control systems of heat and power plants. SIL-requirements are relatively new within the domain and there is a need for guidance to be able to follow the requirements. The target of this report is the people who work with safety questions during new construction, reconstruction, or modification of furnace plants. In the work, the Pressure Equipment Directive, 97/23/EC, as well as standards which use the SIL concept have been studied. Additionally, standards for water-tube boilers have been studied. The focus has been on the safety systems (safety functions) which are used in water-tube boilers for heat and power plants; other systems, which are parts of these boilers, have not been considered. Guidance has been given for the aforementioned standards as well as safety requirements specification and risk analysis. An old hot water generator and a relatively new steam boiler have been used as case studies. The design principles and safety functions of the furnaces have been described. During the risk analysis important hazards were identified. A method for performing a risk analysis has been described and the appropriate content of a safety requirements specification has been defined. If a heat or power plant is constructed, modified, or reconstructed, a safety life cycle shall be followed. The purpose of the safety life cycle is to plan, describe, document, perform, check, test, and validate that everything is correctly done. The components of the safety

  6. Correct safety requirements during the life cycle of heating plants; Korrekta saekerhetskrav under vaermeanlaeggningars livscykel

    Energy Technology Data Exchange (ETDEWEB)

    Tegehall, Jan; Hedberg, Johan [Swedish National Testing and Research Inst., Boraas (Sweden)

    2006-10-15

    The safety of old steam boilers or hot water generators is in principle based on electromechanical components which are generally easy to understand. The use of safety-PLC is a new and flexible way to design a safe system. A programmable system offers more degrees of freedom and consequently new problems may arise. As a result, new standards which use the Safety Integrity Level (SIL) concept for the level of safety have been elaborated. The goal is to define a way of working to handle requirements on safety in control systems of heat and power plants. SIL-requirements are relatively new within the domain and there is a need for guidance to be able to follow the requirements. The target of this report is the people who work with safety questions during new construction, reconstruction, or modification of furnace plants. In the work, the Pressure Equipment Directive, 97/23/EC, as well as standards which use the SIL concept have been studied. Additionally, standards for water-tube boilers have been studied. The focus has been on the safety systems (safety functions) which are used in water-tube boilers for heat and power plants; other systems, which are parts of these boilers, have not been considered. Guidance has been given for the aforementioned standards as well as safety requirements specification and risk analysis. An old hot water generator and a relatively new steam boiler have been used as case studies. The design principles and safety functions of the furnaces have been described. During the risk analysis important hazards were identified. A method for performing a risk analysis has been described and the appropriate content of a safety requirements specification has been defined. If a heat or power plant is constructed, modified, or reconstructed, a safety life cycle shall be followed. The purpose of the safety life cycle is to plan, describe, document, perform, check, test, and validate that everything is correctly done. The components of the safety

  7. Pilot plants for polymers: Safety considerations

    Energy Technology Data Exchange (ETDEWEB)

    Cordeiro, C.F.; Zvanut, C.W.

    1986-01-01

    Air Products and Chemicals is a major manufacturer of polyvinyl alcohol, vinyl acetate-ethylene emulsions and suspension PVC. Polyvinyl alcohol is a water soluble polymer and its primary end-uses are as a textile sizing agent and in adhesives. The emulsion products are used primarily in adhesives, paper, paints, and non-wovens. In order to support these business areas and to expand into new product lines, Air Products operates several polymer pilot plants. The safe operation of these pilot plants mandates careful attention to both design and operating procedures. Often, more care is needed in operating a polymer pilot plant than in other pilot plants or manufacturing facilities.

  8. Woody plants and woody plant management: ecology, safety, environmental impact

    Science.gov (United States)

    James H. Miller

    2001-01-01

    Wise and effective woody plant management is an increasing necessity for many land uses and conservation practices, especially on forests and rangelands where native or exotic plants are affecting productivity, access, or critical habitat. Tools and approaches for managing woody plants have been under concerted development for the past 50 years, integrating mechanical...

  9. Food and feed safety aspects of cisgenic crop plant varieties

    NARCIS (Netherlands)

    Prins, T.W.; Kok, E.J.

    2010-01-01

    This report presents the results of the discussions that identified food and feed safety aspects of cisgenic plant varieties in comparison to conventional varieties on the one hand and transgenic plant varieties on the other hand. It was concluded that on the basis of the general characteristics of

  10. Comparative safety assessment of plant-derived foods.

    Science.gov (United States)

    Kok, E J; Keijer, J; Kleter, G A; Kuiper, H A

    2008-02-01

    The second generation of genetically modified (GM) plants that are moving towards the market are characterized by modifications that may be more complex and traits that more often are to the benefit of the consumer. These developments will have implications for the safety assessment of the resulting plant products. In part of the cases the same crop plant can, however, also be obtained by 'conventional' breeding strategies. The breeder will decide on a case-by-case basis what will be the best strategy to reach the set target and whether genetic modification will form part of this strategy. This article discusses important aspects of the safety assessment of complex products derived from newly bred plant varieties obtained by different breeding strategies. On the basis of this overview, we conclude that the current process of the safety evaluation of GM versus conventionally bred plants is not well balanced. GM varieties are elaborately assessed, yet at the same time other crop plants resulting from conventional breeding strategies may warrant further food safety assessment for the benefit of the consumer. We propose to develop a general screening frame for all newly developed plant varieties to select varieties that cannot, on the basis of scientific criteria, be considered as safe as plant varieties that are already on the market.

  11. Accident Safety Analysis Method Study for Spent Fuel Reprocessing Plant%乏燃料后处理厂事故安全分析方法的探讨

    Institute of Scientific and Technical Information of China (English)

    李锐柔; 徐云起

    2012-01-01

    According to some relative documents (like NRC, DOE and IAEA documents et al. ), and considering the experience and practical technology level of the safety analysis of spent fuel reprocess-ing plant in China, this article suggested that the risk assessment, which combining both deterministic ap-proach and probabilistic safety analysis, could be applied to the accident safety analysis of spent fuel repro-cessing plant in China. Meanwhile, the coordinated working procedure was also proposed.%参考了NRC、DOE、IAEA等相关文件,结合我国乏燃料后处理厂安全分析的经验和实际技术水平,建议我国乏燃料后处理厂在事故安全分析中可采用确定论和概率安全分析相结合的风险评价方法,并提出了相应的工作流程。

  12. Rankine bottoming cycle safety analysis. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lewandowski, G.A.

    1980-02-01

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

  13. Evolution of the future plants operation for a better safety

    Energy Technology Data Exchange (ETDEWEB)

    Papin, B.; Malvache, P.

    1994-12-31

    This paper describes a coordinated research project of the french CEA, addressing to the evolutions in plant operation apt to bring perceptible and assessable improvement in the operational safety. This program has been scheduled for the 1992-1996 period, with a global 40 men/year effort. The present status of the two main parts of the project is presented: ESCRIME (program aiming at defining the optimal share of tasks between humans and computers in plant operation), IMAGIN (research in the domain of plant information management, in order to ensure the global coherence of the image of the plant, used by the different actors in plant operation). (authors). 3 refs., 4 figs.

  14. Safety studies on Korean fusion DEMO plant using Integrated Safety Assessment Methodology: Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

    2015-10-15

    Highlights: • The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. • The concepts of Integrated Safety Assessment Methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. • Phenomena Identification Ranking Table (PIRT) and Objective Provision Tree (OPT) were performed and updated. • This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. - Abstract: The purpose of this paper is to investigate safety issues using Integrated Safety Assessment Methodology (ISAM) proposed by Generation IV Forum Risk and Safety Working Group (RSWG) for Korean fusion DEMO plant (K-DEMO). In ongoing nuclear energy research such as Generation IV fission power plant (GEN-IV), new methodology based on Technology-Neutral Framework (TNF) has been applied for safety assessment. In this methodology, design and regulatory requirements for safety of nuclear power plants are considered simultaneously. The design based on regulatory requirements can save resource, time, and manpower while maintaining high level safety. ISAM is one of the options to apply TNF in K-DEMO. We have performed safety studies for K-DEMO using Phenomena Identification and Ranking Table (PIRT) and Objective Provision Tree (OPT) which are constitutive part of ISAM. Considering the design phase of K-DEMO, the current study focused on PIRT and OPT for K-DEMO. Results have been reviewed and updated by Korean fusion advisory group after considering the views of specialists from domestic universities, industries, and national institutes in South Korea.

  15. Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

  16. Risk Analysis on Safety Injection Test of PWR Nuclear Power Plant%压水堆核电厂安注试验风险研究

    Institute of Scientific and Technical Information of China (English)

    徐永华

    2012-01-01

    During the safety injection (SI) test in the PWR nuclear power plant, full water in pressurizer and high-high level, low-low level in SG may take place. This paper analyzes the risks and response measures in the SI test. Focusing on the full scale problem of pressurizer thermal calibration level gauge in the SI test of nuclear power plant, the test process is analyzed and the problems that should be noted in the SI test are summed up.%在压水堆核电厂安注试验期间,可能出现稳压器满水及主蒸汽发生器(SG)高高、低低液位等问题.本文分析了安注试验的风险及应对措施;针对某核电厂进行安注试验时稳压器热态标定液位计满量程问题,对试验过程进行分析,并总结出安注试验中应注意的问题.

  17. Safety and Radiation Protection at Swedish Nuclear Power Plants 2005

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-15

    In 2005, no severe events occurred which challenged the safety at the Swedish nuclear power plants. However, some events have been given a special focus. The 'Gudrun' storm, which occurred in January 2005, affected the operation of the reactors at Ringhals and Barsebaeck 2. At Ringhals, the switchyards were affected by salt deposits and, at Barsebaeck, the 400kV grid was subjected to interruptions. The long-term trend is that the total number of fuel defects in Swedish reactors is decreasing. The damage that occurs nowadays has mainly been caused by small objects entering the fuel via the coolant and fretting holes in the cladding. To reduce the number of defects of this type, fuel with filters is successively being introduced to prevent debris from entering the fuel assemblies and cyclone filters in the facility which cleans the coolant. Since the mid-nineties, the pressurised water reactors, Ringhals 2, 3 and 4, have had problems with fuel rod bowing in excess of the safety analysis calculations. Ringhals AB (RAB) has adopted measures to rectify the bowing. Follow-up work shows that the fuel rod bowing is decreasing. The followup in 2005 of damaged tubes in the Ringhals 4 steam generators indicates a continued slow damage propagation. Tubes with defects of such a limited extent that there are adequate margins to rupture and loosening have been kept in operation. Damaged tubes with insufficient margins have plugged. During the year, previously observed minor leakage from the reactor containment in Ringhals 2 was investigated in greater detail and repaired. The investigations showed extensive corrosion attack caused by deficiencies in connection with containment construction. The ageing of electrical cables and other equipment in the I-C systems has been examined by SKI. Regulatory supervision has so far shown that these issues are largely handled in a satisfactory manner by the licensees but that certain supplementary investigations and other measures

  18. Improved Management of Part Safety Classification System for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Young; Park, Youn Won; Park, Heung Gyu; Park, Hyo Chan [BEES Inc., Daejeon (Korea, Republic of)

    2016-10-15

    As, in recent years, many quality assurance (QA) related incidents, such as falsely-certified parts and forged documentation, etc., were reported in association with the supply of structures, systems, components and parts to nuclear power plants, a need for a better management of safety classification system was addressed so that it would be based more on the level of parts . Presently, the Korean nuclear power plants do not develop and apply relevant procedures for safety classifications, but rather the safety classes of parts are determined solely based on the experience of equipment designers. So proposed in this paper is a better management plan for safety equipment classification system with an aim to strengthen the quality management for parts. The plan was developed through the analysis of newly introduced technical criteria to be applied to parts of nuclear power plant.

  19. Work practices, fatigue, and nuclear power plant safety performance.

    Science.gov (United States)

    Baker, K; Olson, J; Morisseau, D

    1994-06-01

    This paper focuses on work practices that may contribute to fatigue-induced performance decrements in the commercial nuclear power industry. Specifically, the amount of overtime worked by operations, technical, and maintenance personnel and the 12-h operator shift schedule are studied. Although overtime for all three job categories was fairly high at a number of plants, the analyses detected a clear statistical relationship only between operations overtime and plant safety performance. The results for the 12-h operator shift schedule were ambiguous. Although the 12-h operator shift schedule was correlated with operator error, it was not significantly related to the other five safety indicators. This research suggests that at least one of the existing work practices--the amount of operator overtime worked at some plants--represents a safety concern in this industry; however, further research is required before any definitive conclusions can be drawn.

  20. [Safety assessment of foods derived from genetically modified plants].

    Science.gov (United States)

    Pöting, A; Schauzu, M

    2010-06-01

    The placing of genetically modified plants and derived food on the market falls under Regulation (EC) No. 1829/2003. According to this regulation, applicants need to perform a safety assessment according to the Guidance Document of the Scientific Panel on Genetically Modified Organisms of the European Food Safety Authority (EFSA), which is based on internationally agreed recommendations. This article gives an overview of the underlying legislation as well as the strategy and scientific criteria for the safety assessment, which should generally be based on the concept of substantial equivalence and carried out in relation to an unmodified conventional counterpart. Besides the intended genetic modification, potential unintended changes also have to be assessed with regard to potential adverse effects for the consumer. All genetically modified plants and derived food products, which have been evaluated by EFSA so far, were considered to be as safe as products derived from the respective conventional plants.

  1. Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kisner, Roger A [ORNL; Mullens, James Allen [ORNL; Wilson, Thomas L [ORNL; Wood, Richard Thomas [ORNL; Korsah, Kofi [ORNL; Qualls, A L [ORNL; Muhlheim, Michael David [ORNL; Holcomb, David Eugene [ORNL; Loebl, Andy [ORNL

    2007-08-01

    Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

  2. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Energy Technology Data Exchange (ETDEWEB)

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  3. Periodic safety review of French nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Poirrier, D.; Debes, M. [Electricite de France, Paris (France)

    1997-12-01

    The safety of nuclear power plants (NPPs) is checked through different types of safety evaluations, for example, a continuous process, with followup of operational feedback and over-all evaluation every year by each NPP; specific examination, with the study of generic problems when they occur; and a 10-yr outage inspection. In France, the license does not explicitly require periodic safety reviews (PSRs), but an article has been added to the Decree of December 11, 1963 concerning nuclear installations that states, {open_quotes}The Ministers may jointly request the operating utility at any time to proceed to a review of nuclear safety,{close_quotes} which supports requests for PSRs from the safety authority.

  4. IR-360 nuclear power plant safety functions and component classification

    Energy Technology Data Exchange (ETDEWEB)

    Yousefpour, F., E-mail: fyousefpour@snira.co [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of); Shokri, F.; Soltani, H. [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of)

    2010-10-15

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  5. A proposal of safety indicators aggregation to assess the safety management effectiveness of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Jose Antonio B.; Saldanha, Pedro L.C. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Coordenacao-Geral de Reatores e Ciclo Combustivel], e-mail: jantonio@cnen.gov.br, e-mail: saldanha@cnen.gov.br; Melo, Paulo F.F. Frutuoso e [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: frutuoso@con.ufrj.br

    2009-07-01

    Safety management has changed with the evolution of management methods, named Quality Systems, moving from Quality Control, where the focus was the product, passing through Quality Assurance, which takes care of the whole manufacturing process and reaching the Total Quality Management, where policies and goals are established. Nowadays, there is a trend towards Management Systems, which integrate all different aspects related to the management of an organization (safety, environment, security, quality, costs and, etc), but it is necessary to have features to establish and assure that safety overrides the remaining aspects. The most usual way to reach this goal is to establish a policy where safety is a priority, but its implementation and the assessment of its effectiveness are no so simple. Nuclear power plants usually have over a hundred safety indicators in many processes dedicated to prevent and detect problems, although a lot of them do not evaluate these indicators in an integrated manner or point out degradation trends of organizational aspects, which can affect the plant safety. This work develops an aggregation of proactive and reactive safety indicators in order to evaluate the effectiveness of nuclear power plant safety management and to detect, at early stages, signs of process degradation or activities used to establish, maintain and assure safety conditions. The aggregation integrates indicators of the usual processes and is based on the manner the management activities have been developed in the last decades, that is: Planning, Doing, Checking and Acting - known as PDCA cycle - plus a fifth element related to the capability of those who perform safety activities. The proposed aggregation is in accordance to Brazilian standards and international recommendations and constitutes a friendly link between the top management level and the daily aspects of the organization. (author)

  6. Uncertainty Quantification for Safety Verification Applications in Nuclear Power Plants

    Science.gov (United States)

    Boafo, Emmanuel

    There is an increasing interest in computational reactor safety analysis to systematically replace the conservative calculations by best estimate calculations augmented by quantitative uncertainty analysis methods. This has been necessitated by recent regulatory requirements that have permitted the use of such methods in reactor safety analysis. Stochastic uncertainty quantification methods have shown great promise, as they are better suited to capture the complexities in real engineering problems. This study proposes a framework for performing uncertainty quantification based on the stochastic approach, which can be applied to enhance safety analysis. (Abstract shortened by ProQuest.).

  7. 优化设计输入和分析方法以提高核电厂抗震安全性%Optimize the design input and analysis methods to improve the seismic safety of nuclear power plants

    Institute of Scientific and Technical Information of China (English)

    张超琦; 杨建华

    2013-01-01

      日本福岛事故后核电厂抵御极端自然灾害能力受到广泛关注,世界上各个国家都积极开展相关研究,而地震一直是核电厂工程安全问题的主要威胁之一,因此核电厂的抗震安全性更是成为业界分析研究的重点。楼层反应谱作为核电厂系统、结构和部件抗震设计的输入,其计算分析是核电厂抗震分析的重要环节,其结果对于核电厂的抗震安全水平起着举足轻重的作用。本文以中国核电工程有限公司自主研发的三代机型ACP1000标准设计为例,通过介绍楼层反应谱的输入、分析过程和方法,来阐述合理确定符合国情的地震输入、采用先进的建模和分析方法,对完善核电厂的抗震设计、提高核电厂的抗震安全性具有重要意义。%The competence of resisting nature extreme disaster is widely concerned after Fukushima nu-clear incident in Japan. Many countries carry through investigation actively. The seismic is one of primary threat-ens for nuclear power plant safety. Therefore the seismic ability of nuclear power plants becomes the investiga-tive emphasis. The floor response spectra is the input of the seismic design for nuclear plant systems,structures and components,and the important part of the seismic design,which is holding the balance in nuclear plant seis-mic safety. In this paper we use the third generation nuclear power plant example that is ACP1000 normal design excogitated by the China Nuclear Power Engineering Co.,LTD. themselves to expatiate the reasonable seismic input and the advanced analysis method for China. Through introducing the input,the analysis process and the method about the ACP1000 floor response spectra are calculated. Then it has significant effect for improving the nuclear power plant seismic design and seismic safety.

  8. The application of transcriptomics in the comparative safety assessment of (GMO-derived) plant products

    OpenAIRE

    Kok, E.J.

    2008-01-01

    National and international organizations have discussed current approaches to the safety assessment of complex (plant) food products in general and the safety assessment of GMO-derived food products in particular. One of the recommendations of different expert meetings was that the new analytical techniques, in particular the ‘omics’ approaches, need to be explored for their potential to improve the analysis and thereby the toxicological and nutritional assessment of complex (GMO-derived) pla...

  9. 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.

  10. Hydrogen energy demonstration plant in Patagonia: Description and safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Aprea, Jose Luis [CNEA (Argentine Atomic Energy Commission), AAH, IRAM, Comahue University, CC 805, 8300 Neuquen (Argentina)

    2009-05-15

    Hydrogen safety issues and especially hydrogen hazard's address are key points to remove any safety-related barrier in the implementation process of hydrogen energy systems. Demonstrative systems based on hydrogen technologies represent a clear contribution to the task of showing the feasibility of the new technologies and their beneficial capabilities among the public. In this paper, the safety features of the first hydrogen energy demonstrative plant conceived in Latin America are analyzed. The facilities, located in the village of Pico Truncado, Patagonia, Argentina, serve to gain day-to-day experience in the production, storage, distribution, conversion and use of hydrogen in several applications. The plant uses electrolysis to generate pure hydrogen from renewable primary sources, taking advantage of the installed wind power capacity that is continually growing in the region. The installations were designed to accomplish with two primary objectives: total safety assurance and minimization of human errors. Some details of the plant, including a general layout, are presented here, in addition with design criteria, hydrogen hazards, structural precautions, gas monitoring system, existing regulations and safety requirements. (author)

  11. South Ukraine NPP: Safety improvements through Plant Computer upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Brenman, O. [Westinghouse Electric Company, 4350 Northern Pike, Monroeville, PA 15146 (United States); Chernyshov, M. A. [Westron, LLC, 1 Acad. Proskura St., Kharkiv 61070 (Ukraine); Denning, R. S. [Battelle, 505 King Ave, Columbus, OH 43201 (United States); Kolesov, S. A. [NAEK Energoatom, 3 Vetrov Str., Kiev, 01032 (Ukraine); Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I. [PO South Ukraine NPP, NAEK Energoatom, Mylolayv Region, 55000 (Ukraine); Trosman, G. [US Dept. of Energy, International Nuclear Safety Program, Washington, DC 20585 (United States)

    2006-07-01

    This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)

  12. Nuclear power plant safety related pump issues

    Energy Technology Data Exchange (ETDEWEB)

    Colaccino, J.

    1996-12-01

    This paper summarizes of a number of pump issues raised since the Third NRC/ASME Symposium on Valve and Pump Testing in 1994. General issues discussed include revision of NRC Inspection Procedure 73756, issuance of NRC Information Notice 95-08 on ultrasonic flow meter uncertainties, relief requests for tests that are determined by the licensee to be impractical, and items in the ASME OM-1995 Code, Subsection ISTB, for pumps. The paper also discusses current pump vibration issues encountered in relief requests and plant inspections - which include smooth running pumps, absolute vibration limits, and vertical centrifugal pump vibration measurement requirements. Two pump scope issues involving boiling water reactor waterlog and reactor core isolation cooling pumps are also discussed. Where appropriate, NRC guidance is discussed.

  13. Comparative safety assessment of plant-derived foods

    NARCIS (Netherlands)

    Kok, E.J.; Keijer, J.; Kleter, G.A.; Kuiper, H.A.

    2008-01-01

    The second generation of genetically modified (GM) plants that are moving towards the market are characterized by modifications that may be more complex and traits that more often are to the benefit of the consumer. These developments will have implications for the safety assessment of the resulting

  14. 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)

  15. Manpower analysis in transportation safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-05-01

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

  16. HFE safety reviews of advanced nuclear power plant control rooms

    Science.gov (United States)

    Ohara, John

    1994-01-01

    Advanced control rooms (ACR's) will utilize human-system interface (HSI) technologies that may have significant implications for plant safety in that they will affect the operator's overall role and means of interacting with the system. The Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) aspects of HSI's to ensure that they are designed to good HFE principles and support performance and reliability in order to protect public health and safety. However, the only available NRC guidance was developed more than ten years ago, and does not adequately address the human performance issues and technology changes associated with ACR's. Accordingly, a new approach to ACR safety reviews was developed based upon the concept of 'convergent validity'. This approach to ACR safety reviews is described.

  17. Safety culture assessment in petrochemical industry: a comparative study of two algerian plants.

    Science.gov (United States)

    Boughaba, Assia; Hassane, Chabane; Roukia, Ouddai

    2014-06-01

    To elucidate the relationship between safety culture maturity and safety performance of a particular company. To identify the factors that contribute to a safety culture, a survey questionnaire was created based mainly on the studies of Fernández-Muñiz et al. The survey was randomly distributed to 1000 employees of two oil companies and realized a rate of valid answer of 51%. Minitab 16 software was used and diverse tests, including the descriptive statistical analysis, factor analysis, reliability analysis, mean analysis, and correlation, were used for the analysis of data. Ten factors were extracted using the analysis of factor to represent safety culture and safety performance. The results of this study showed that the managers' commitment, training, incentives, communication, and employee involvement are the priority domains on which it is necessary to stress the effort of improvement, where they had all the descriptive average values lower than 3.0 at the level of Company B. Furthermore, the results also showed that the safety culture influences the safety performance of the company. Therefore, Company A with a good safety culture (the descriptive average values more than 4.0), is more successful than Company B in terms of accident rates. The comparison between the two petrochemical plants of the group Sonatrach confirms these results in which Company A, the managers of which are English and Norwegian, distinguishes itself by the maturity of their safety culture has significantly higher evaluations than the company B, who is constituted of Algerian staff, in terms of safety management practices and safety performance.

  18. 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.

  19. Efficacy and safety of plant stanols and sterols in the control of blood cholesterol levels

    NARCIS (Netherlands)

    Katan, M.B.; Grundy, S.M.; Jones, P.J.H.; Law, M.R.; Miettinen, T.; Paoletti, R.

    2003-01-01

    Foods with plant stanol or sterol esters lower serum cholesterol levels. We summarize the deliberations of 32 experts on the efficacy and safety of sterols and stanols. A meta-analysis of 41 trials showed that intake of 2 g/d of stanols or sterols reduced low-density lipoprotein (LDL) by 10%; higher

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

    Energy Technology Data Exchange (ETDEWEB)

    Menzel, Francine; Sabundjian, Gaiane, E-mail: fmenzel@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); D' auria, Francesco, E-mail: f.dauria@ing.unipi.it [Universita degli Studi di Pisa, Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG), Pisa (Italy); Madeira, Alzira A., E-mail: alzira@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    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)

  1. SMV model-based safety analysis of software requirements

    Energy Technology Data Exchange (ETDEWEB)

    Koh, Kwang Yong [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Seong, Poong Hyun [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)], E-mail: phseong@kaist.ac.kr

    2009-02-15

    Fault tree analysis (FTA) is one of the most frequently applied safety analysis techniques when developing safety-critical industrial systems such as software-based emergency shutdown systems of nuclear power plants and has been used for safety analysis of software requirements in the nuclear industry. However, the conventional method for safety analysis of software requirements has several problems in terms of correctness and efficiency; the fault tree generated from natural language specifications may contain flaws or errors while the manual work of safety verification is very labor-intensive and time-consuming. In this paper, we propose a new approach to resolve problems of the conventional method; we generate a fault tree from a symbolic model verifier (SMV) model, not from natural language specifications, and verify safety properties automatically, not manually, by a model checker SMV. To demonstrate the feasibility of this approach, we applied it to shutdown system 2 (SDS2) of Wolsong nuclear power plant (NPP). In spite of subtle ambiguities present in the approach, the results of this case study demonstrate its overall feasibility and effectiveness.

  2. 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.

  3. Different sets of reliability data and success criteria in a probabilistic safety assessment for a plant producing nitroglycol.

    Science.gov (United States)

    Hauptmanns, Ulrich

    2009-03-15

    The lack of plant-specific reliability data for probabilistic safety assessments usually makes it necessary to use generic reliability data. Justifiably different assessments of plant behaviour (success criteria) lead to different models of plant systems. Both affect the numerical results of a probabilistic safety assessment. It is shown how these results change, if different sets of reliability data and different choices of success criteria for the safety system are employed. Differences in results may influence decisions taken on their basis and become especially important if compliance with a safety goal has to be proved, e.g. a safety integrity level. For the purpose of demonstration an accident sequence from a probabilistic safety assessment of a plant producing nitroglycol is used. The analysis relies on plant-specific reliability data so that it provides a good yardstick for comparing it with results obtained using generic data. The superiority of plant-specific data, which should of course be acquired, cannot be doubted. Nevertheless, plant safety can be improved even if generic data are used. However, the assignment to a safety integrity level may be affected by differences in both data and success criteria.

  4. 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.

  5. Personnel Safety for Future Magnetic Fusion Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee Cadwallader

    2009-07-01

    The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

  6. 天然气净化厂安全防火防爆系统的应用分析%Application and analysis of the safety and fire protection and explosion protection system of natural gas purification plant

    Institute of Scientific and Technical Information of China (English)

    都永昌

    2016-01-01

    本文就天然气净化厂安全防火防爆系统的应用进行分析,以促进天然气净化厂对安全工作的重视程度,提升安全管理水平。%In this paper, the application of natural gas purification plant safety fire and explosion protection system is analyzed, in order to promote the natural gas purification plant to the safety work of attention, improve the level of safety management.

  7. Technical Safety Appraisal of the Rocky Flats Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Blake P.

    1989-01-01

    This report provides the results of a Technical Safety Appraisal (TSA) of the Rocky Flats Plant (RFP) conducted November 14 to 18 and November 28 to December 9, 1988. This appraisal covered the effectiveness and improvements in the RFP safety program across the site, evaluating progress to date against standards of accepted practice. The appraisal included coverage of the timeliness and effectiveness of actions taken in response to the recommendations/concerns in three previous Technical Safety Appraisals (TSAs) of RFP Bldg. 707 conducted in July 1986, Bldgs. 771/774 conducted in October/November 1986, and Bldgs. 776/777 conducted in January/February 1988. Results of this appraisal are given in Section IV for each of 14 technical safety areas at RFP. These results include a discussion, conclusions and any new safety concerns for each technical safety area. Appendix A contains a description of the system for categorizing concerns, and the concerns are tabulated in Appendix B. Appendix C reports on the evaluation of the contractor's actions and the current status of each of the 230 recommendations and concerns contained in the three previous TSA reports.

  8. [Design of a Hazard Analysis and Critical Control Points (HACCP) plan to assure the safety of a bologna product produced by a meat processing plant].

    Science.gov (United States)

    Bou Rached, Lizet; Ascanio, Norelis; Hernández, Pilar

    2004-03-01

    The Hazard Analysis and Critical Control Point (HACCP) is a systematic integral program used to identify and estimate the hazards (microbiological, chemical and physical) and the risks generated during the primary production, processing, storage, distribution, expense and consumption of foods. To establish a program of HACCP has advantages, being some of them: to emphasize more in the prevention than in the detection, to diminish the costs, to minimize the risk of manufacturing faulty products, to allow bigger trust to the management, to strengthen the national and international competitiveness, among others. The present work is a proposal based on the design of an HACCP program to guarantee the safety of the Bologna Special Type elaborated by a meat products industry, through the determination of hazards (microbiological, chemical or physical), the identification of critical control points (CCP), the establishment of critical limits, plan corrective actions and the establishment of documentation and verification procedures. The used methodology was based in the application of the seven basic principles settled down by the Codex Alimentarius, obtaining the design of this program. In view of the fact that recently the meat products are linked with pathogens like E. coli O157:H7 and Listeria monocytogenes, these were contemplated as microbiological hazard for the establishment of the HACCP plan whose application will guarantee the obtaining of a safe product.

  9. Preliminary Integrated Safety Analysis Status Report

    Energy Technology Data Exchange (ETDEWEB)

    D. Gwyn

    2001-04-01

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

  10. Design data and safety features of commerical nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Heddleson, F.A.

    1976-06-01

    Design data, safety features, and site characteristics are summarized for 34 nuclear power units in 17 power stations in the United States. Six pages of data are presented for each plant, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. This volume covers Light Water Reactors (LWRs) with dockets 50-508 through 50-549, four HTGRs--50-171, 50-267, 50-450/451, 50-463/464, the Atlantic Floating Station 50-477/478, and the Clinch River Breeder 50-537.

  11. 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.

  12. Considerations on Safety Evaluation of Safety grade Smart Transmitter in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Tae; Jeong, Choong heui [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    Non-safety grade smart transmitters have been used for I and C systems of NPPs(Nuclear Power Plants). Smart transmitter is a microprocessor-based device including software and provides capability for digital signals to be communicated on top of the 4-20 mA analog signals. Recently, smart transmitters have been used for safety grade I and C systems as well as non-safety grade I and C system for SKN 3 and 4. Due to potential benefits of smart transmitter, it is anticipated smart transmitters will be widely used safety-related applications at NPPs. For those reasons, smart transmitter's technology and characteristics need to be investigated. Smart transmitters have been used for safety grade as well as non-safety grade I and C system since SKN 3 and 4. Due to potential benefits of smart transmitter, it is anticipated smart transmitters will be widely used safety-related applications at NPPs. For those reasons, smart transmitter's technology and characteristics need to be investigated. To get useful information about that, we surveyed EPRI qualification report, NRC event report, and SKN 3 and 4's review.

  13. Safety Evaluation for Packaging (onsite) T Plant Canyon Items

    Energy Technology Data Exchange (ETDEWEB)

    OBRIEN, J.H.

    2000-07-14

    This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments.

  14. 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.

  15. 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...

  16. Plants and parts of plants used in food supplements: an approach to their safety assessment

    Directory of Open Access Journals (Sweden)

    Brunella Carratù

    2010-12-01

    Full Text Available In Italy most herbal products are sold as food supplements and are subject only to food law. A list of about 1200 plants authorised for use in food supplements has been compiled by the Italian ministry of Health. In order to review and possibly improve the ministry's list an ad hoc working group of Istituto Superiore di Sanità was requested to provide a technical and scientific opinion on plant safety. The listed plants were evaluated on the basis of their use in food, therapeutic activity, human toxicity and in no-alimentary fields. Toxicity was also assessed and plant limitations to use in food supplements were defined.

  17. Updated safety analysis of ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  18. Preliminary safety analysis for key design features of KALIMER-600

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Chang, W. P.; Suk, S. D.; Ha, K. S.; Jeong, H. Y.; Heo, S

    2004-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents in the KALIMER design with breakeven core are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2. In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign 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 using the SSC-K code to investigate the KALIMER-600 system response to the events. They are categorized as Bounding Events (BEs) because of their low probability of occurrence. In Chapter 4, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed.The performance analysis of the KALIMER-600 containment and some evaluations for the behaviors during HCDA will be performed later.

  19. Factors influencing workers to follow food safety management systems in meat plants in Ontario, Canada.

    Science.gov (United States)

    Ball, Brita; Wilcock, Anne; Aung, May

    2009-06-01

    Small and medium sized food businesses have been slow to adopt food safety management systems (FSMSs) such as good manufacturing practices and Hazard Analysis Critical Control Point (HACCP). This study identifies factors influencing workers in their implementation of food safety practices in small and medium meat processing establishments in Ontario, Canada. A qualitative approach was used to explore in-plant factors that influence the implementation of FSMSs. Thirteen in-depth interviews in five meat plants and two focus group interviews were conducted. These generated 219 pages of verbatim transcripts which were analysed using NVivo 7 software. Main themes identified in the data related to production systems, organisational characteristics and employee characteristics. A socio-psychological model based on the theory of planned behaviour is proposed to describe how these themes and underlying sub-themes relate to FSMS implementation. Addressing the various factors that influence production workers is expected to enhance FSMS implementation and increase food safety.

  20. Safety Management Characteristics Reflected in Interviews at Swedish Nuclear Power Plants: A System Perspective Approach

    Energy Technology Data Exchange (ETDEWEB)

    Salo, Ilkka (Risk Analysis, Social and Decision Research Unit, Dept. of Psychology, Stockholm Univ., Stockholm (Sweden))

    2005-12-15

    The present study investigated safety management characteristics reflected in interviews with participants from two Swedish nuclear power plants. A document analysis regarding the plants' organization, safety policies, and safety culture work was carried out as well. The participants (n=9) were all nuclear power professionals, and the majority managers at different levels with at least 10 years of nuclear power experience. The interview comprised themes relevant for organizational safety and safety management, such as: organizational structures and organizational change, threats to safety, information feedback and knowledge transfer, safety analysis, safety policy, and accident and incident analysis and reporting. The results were in part modeled to important themes derived from a general system theoretical framework suggested by Svenson and developed by Svenson and Salo in relation to studies of 'non-nuclear' safety organizations. A primer to important features of the system theoretical framework is presented in the introductory chapter. The results from the interviews generated interesting descriptions about nuclear safety management in relation to the above themes. Regarding organizational restructuring, mainly centralizations of resources, several examples of reasons for the restructuring and related benefits for this centralization of resources were identified. A number of important reminders that ought to be considered in relation to reorganization were also identified. Regarding threats to the own organization a number of such was interpreted from the interviews. Among them are risks related to generation and competence change-over and risks related to outsourcing of activities. A thorough picture of information management and practical implications related to this was revealed in the interviews. Related to information feedback is the issue of organizational safety indicators and safety indicators in general. The interview answers indicated

  1. 核电厂DCS安全级应用软件开发的危险分析%Hazard analysis of application software development for nuclear power plant DCS safety system

    Institute of Scientific and Technical Information of China (English)

    艾九斤; 李运坚; 李相建

    2012-01-01

    为了减小或避免因控制系统软件而导致的核电厂安全性降低的不良后果,提出了对核电厂数字控制系统安全级应用软件开发过程进行危险分析的活动.采用验证和确认的方法,并结合安全保护层模型、预先危险分析方法(PHA)、故障树分析等方法对应用软件开发过程中的系统设计、软件设计、软件实现各个阶段的危险进行分析.通过CPR1000项目工程实践表明,采用验证和确认的方法能有效地减小软件开发过程中的危险以提高应用软件的安全性,从而最终提高核电厂的安全性.%In order to reduce or avoid the bad consequences of nuclear power plant security reduction caused by the control system software, the hazard analysis activity for the application software development process of nuclear power plant digital control system is put forward. The verification and validation method combined with the safety protection layer model. the preliminary hazard analysis, the event tree analysis model and so on is used to analyze the hazards of application software development process during the system design, software design and software realization phases. The practice of the CPR1000 project indicate that the verification and validation method can effectively reduce the hazards of software development process to enhance the security of the application software, finally the security of the nuclear power plant is enhanced.

  2. Reliability study: digital engineered safety feature actuation system of Korean Standard Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sudarno [National Nuclear Energy Agency, Batan (Indonesia); Kang, H. G.; Jang, S. C.; Eom, H. S.; Ha, J. J. [KAERI, Taejon (Korea, Republic of)

    2003-04-01

    The usage of digital Instrumentation and Control (I and C) in a nuclear power plant becomes more extensive, including safety related systems. The PSA application of these new designs are very important in order to evaluate their reliability. In particular, Korean Standard Nuclear Power Plants (KSNPPs), typically Ulchin 5 and 6 (UCN 5 and 6) reactor units, adopted the digital safety-critical systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS). In this research, we developed fault tree models for assessing the unavailability of the DESFAS functions. We also performed an analysis of the quantification results. The unavailability results of different DESFAS functions showed that their values are comprised from 5.461E-5 to 3.14E-4. The system unavailability of DESFAS AFAS-1 is estimated as 5.461E-5, which is about 27% less than that of analog system if we consider the difference of human failure probability estimation between both analyses. The results of this study could be utilized in risk-effect analysis of KSNPP. We expect that the safety analysis result will contribute to design feedback.

  3. Safety and Radiation Protection at Swedish Nuclear Power Plants 2007

    Energy Technology Data Exchange (ETDEWEB)

    2008-07-01

    The safety level of the plants is maintained at an acceptable level. SKI has in its regulatory supervision not found any known deficiencies in the barriers which could result in release of radioactive substances in excess of the permitted levels. SKI considers that improvements have been implemented during the year in the management, control and following up of safety work at the plants. In some cases, SKI has imposed requirements that improvements be made. Extensive measures are under way at the nuclear power plants to comply with the safety requirements in SKI's regulations, SKIFS 2004:2 concerning the design and construction of nuclear power reactors, and the stricter requirements regarding physical protection. Concurrently preparations are underway at eight of the ten units for thermal power increases. At the Forsmark plant considerable efforts have been during the year to correct the deficiencies in the safety culture and quality assurance system that became apparent in 2006. A programme to improve the execution of activities has been established in accordance with SKI's decision. SKI considers that the plant has developed in a positive direction but that there are further possibilities for improvement with regard to internal control. This is amongst other things concerns the areas internal auditing, independent safety review function, and working methods. SKI has had special supervision of the plant since 28 September, 2006. At the Oskarshamn plant work has been carried out to improve the organisation and routines in several areas. The plant has established routines which provide the basis to ensure that decisions are taken in a stringent manner. The quality assurance system has a clearer structure and there is a better defined division of work. Some measures remain to be dealt with in 2008. The Ringhals plant has also worked with attitudes to routines and internal control. SKI considers that the measures have good prerequisites to provide a

  4. Plutonium Finishing Plant (PFP) Safety Class and Safety Significant Commercial Grade Items (CGI) Critical Characteristic

    Energy Technology Data Exchange (ETDEWEB)

    THOMAS, R.J.

    2000-04-24

    This document specifies the critical characteristics for Commercial Grade Items (CGI) procured for use in the Plutonium Finishing Plant as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to properly perform its safety function. There may be several manufacturers or models that meet the critical characteristics of any one item.

  5. 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.

  6. Fuzzy-logic-based safety verification framework for nuclear power plants.

    Science.gov (United States)

    Rastogi, Achint; Gabbar, Hossam A

    2013-06-01

    This article presents a practical implementation of a safety verification framework for nuclear power plants (NPPs) based on fuzzy logic where hazard scenarios are identified in view of safety and control limits in different plant process values. Risk is estimated quantitatively and compared with safety limits in real time so that safety verification can be achieved. Fuzzy logic is used to define safety rules that map hazard condition with required safety protection in view of risk estimate. Case studies are analyzed from NPP to realize the proposed real-time safety verification framework. An automated system is developed to demonstrate the safety limit for different hazard scenarios.

  7. Development of Safety Analysis Technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Kwon, Y. M.; Kim, E. K. [KAERI, Daejeon (Korea, Republic of)

    2007-06-15

    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

  8. Unavailability analysis of digital engineered safety feature actuation system

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun Gook; Jang, Seung Cheol [KAERI, Taejon (Korea, Republic of); Wiharjo, Sudarno [National Nuclear Agency-BATAN, Tangerang (Indonesia)

    2003-07-01

    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. Similarly to the PSA result of Digital Plant Protection System, we find out some factors 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.

  9. Plant natural variability may affect safety assessment data.

    Science.gov (United States)

    Batista, Rita; Oliveira, Margarida

    2010-12-01

    Before market introduction, genetic engineered (GE) food products, like any other novel food product, are subjected to extensive assessment of their potential effects on human health. In recent years, a number of profiling technologies have been explored aiming to increase the probability of detecting any unpredictable unintended effect and, consequently improving the efficiency of GE food safety assessment. These techniques still present limitations associated with the interpretation of the observed differences with respect to their biological relevance and toxicological significance. In order to address this issue, in this study, we have performed 2D-gel electrophoresis of five different ears of five different MON810 maize plants and of other five of the non-transgenic near-isogenic line. We have also performed 2D-gel electrophoresis of the pool of the five protein extractions of MON810 and control lines. We have notice that, in this example, the exclusive use of data from 2D-electrophoresed pooled samples, to compare these two lines, would be insufficient for an adequate safety evaluation. We conclude that, when using "omics" technologies, it is extremely important to eliminate all potential differences due to factors not related to the ones under study, and to understand the role of natural plant-to-plant variability in the encountered differences.

  10. 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

  11. 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

  12. Analysis and Application of Nuclear Plant Safety Digital Control System Design Criteria%核电站安全级数字化仪控系统设计准则的分析与应用

    Institute of Scientific and Technical Information of China (English)

    姚光霖; 孙武

    2015-01-01

    Based on the China's nuclear power plant safety digital control system design criteria that the regulations and standards requirements, Analysis of the relationship between the design criteria and its practica-bility is presented.The design criteria are generally recapitulative and don't specify specific regulations or meth-ods, so the detailed methods which satisfy design criteria are established by comparison analysis and application of the design criteria.Through detailed study and application analysis of the design criteria based on Triconex platform for safety-class I&C system, the rationality of established methods are verified.%基于我国核电站安全级数字化仪控系统设计中采用的法规、标准所要求的设计准则,分析各设计准则之间的相互关系及可实施性;设计准则通常概括性强且未给出具体的实施细则或方法,通过对比分析各设计准则的具体要求以及它们在实践中的应用,归纳出更加细化的满足设计准则的实现方法。并通过对主要设计准则在基于Triconex平台的安全级数字化仪控系统中的详细研究和应用分析,验证本文归纳出的实现方法的合理性。

  13. 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.

  14. 78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Science.gov (United States)

    2013-05-01

    ... COMMISSION Qualification Tests for Safety-Related Actuators in Nuclear Power Plants AGENCY: Nuclear..., ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants.'' DG-1235 is proposed Revision 1 of RG... Stations in order to demonstrate their ability to perform their intended safety functions under...

  15. Safety analysis of autonomous excavator functionality

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  16. Core safety of Indian nuclear power plants (NPPs) under extreme conditions

    Indian Academy of Sciences (India)

    J B Joshi; A K Nayak; M Singhal; D Mukhopadhaya

    2013-10-01

    Nuclear power is currently the fourth largest source of electricity production in India after thermal, hydro and renewable sources of electricity. Currently, India has 20 nuclear reactors in operation and seven other reactors are under construction. Most of these reactors are indigenously designed and built Heavy Water Reactors. In addition, a 300 MWe Advanced Heavy Water Reactor has already been designed and in the process of deployment in near future for demonstration of power production from Thorium apart from enhanced safety features by passive means. India has ambitious plans to enhance the share of electricity production from nuclear. The recent Fukushima accident has raised concerns of safety of Nuclear Power Plants worldwide. The Fukushima accident was caused by extreme events, i.e., large earthquake followed by gigantic Tsunami which are not expected to hit India’s coast considering the geography of India and historical records. Nevertheless, systematic investigations have been conducted by nuclear scientists in India to evaluate the safety of the current Nuclear Power Plants in case of occurrence of such extreme events in any nuclear site. This paper gives a brief outline of the safety features of Indian Heavy Water Reactors for prevention and mitigation of such extreme events. The probabilistic safety analysis revealed that the risk from Indian Heavy Water Reactors are negligibly small.

  17. 78 FR 67206 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Science.gov (United States)

    2013-11-08

    ... COMMISSION Qualification Tests for Safety-Related Actuators in Nuclear Power Plants AGENCY: Nuclear...-Related Actuators in Nuclear Power Plants.'' This RG is being revised to provide applicants and licensees with the most current information on testing safety-related actuators in nuclear power plants. This...

  18. 76 FR 35861 - Safety Culture at the Waste Treatment and Immobilization Plant

    Science.gov (United States)

    2011-06-20

    ... the Waste Treatment and Immobilization Plant AGENCY: Defense Nuclear Facilities Safety Board. ACTION... Treatment and Immobilization Plant located at the Hanford site in the state of Washington. DATES: Comments... Safety Culture at the Waste Treatment and Immobilization Plant Pursuant to 42 U.S.C. Sec....

  19. 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... safety function, affected processes, cause of the failure, whether the failure was in the context of the... conclusion of each failure investigation of an item relied on for safety or management measure. (b)...

  20. Application of Software Safety Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, G. Y.; Hur, S.; Cheon, S. W.; Kim, D. H.; Lee, D. Y.; Kwon, K. C. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, S. J.; Koo, Y. H. [Doosan Heavy Industries and Construction Co., Daejeon (Korea, Republic of)

    2009-05-15

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

  1. 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...

  2. Thermal Power Plant Performance Analysis

    CERN Document Server

    2012-01-01

    The analysis of the reliability and availability of power plants is frequently based on simple indexes that do not take into account the criticality of some failures used for availability analysis. This criticality should be evaluated based on concepts of reliability which consider the effect of a component failure on the performance of the entire plant. System reliability analysis tools provide a root-cause analysis leading to the improvement of the plant maintenance plan.   Taking in view that the power plant performance can be evaluated not only based on  thermodynamic related indexes, such as heat-rate, Thermal Power Plant Performance Analysis focuses on the presentation of reliability-based tools used to define performance of complex systems and introduces the basic concepts of reliability, maintainability and risk analysis aiming at their application as tools for power plant performance improvement, including: ·         selection of critical equipment and components, ·         defini...

  3. Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program

    Energy Technology Data Exchange (ETDEWEB)

    Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.

    1983-12-01

    The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.

  4. In-plant safety/relief valve discharge load test, Monticello Plant. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Buzek, E.A. (comp.)

    1977-08-01

    This document reports the results of the test program of safety/relief valve (SRV) discharge load phenomena through a ramshead discharge device, and the effects upon the Mark I primary containment torus structure of the Monticello Nuclear Power Plant. The objectives were to provide a data base for verifying/improving bubble pressure, water reflood and piping load analytical models, and to measure the structural response of the torus, SRV piping, supports and acceleration of the basement and pedestal.

  5. Implementation of external hazards in Probabilistic Safety Assessment for nuclear power plants

    Science.gov (United States)

    Kumar, Manorma; Klug, Joakim; Raimond, Emmanuel

    2015-04-01

    The paper will focus on the discussion on implementation of external hazards in the probabilistic safety assessment (PSA) methods for the extreme external hazards mainly focused on Seismic, Flooding, Meteorological Hazards (e.g. Storm, Extreme temperature, snow pack), Biological infestation, Lightening hazards, Accidental Aircraft crash and man- made hazards including natural external fire and external explosion. This will include discussion on identification of some good practices on the implementation of external hazards in Level 1 PSA, with a perspective of development of extended PSA and introduction of relevant modelling for external hazards in an existing Level 1 PSA. This paper is associated to the European project ASAMPSAE (www.asampsa.eu) which gathers more than 30 organizations (industry, research, safety control) from Europe, US and Japan and which aims at identifying some meaningful practices to extend the scope and the quality of the existing probabilistic safety analysis developed for nuclear power plants.

  6. Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

    Science.gov (United States)

    Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

    2015-04-01

    -tree procedure. Earlier studies have shown that the potentially liquefiable layer at Paks Nuclear Power Plant is situated in relatively large depth. Therefore the applicability and adequacy of the methods at high overburden pressure is important. In case of existing facilities, the geotechnical data gained before construction aren't sufficient for the comprehensive liquefaction analysis. Performance of new geotechnical survey is limited. Consequently, the availability of the data has to be accounted while selection the analysis methods. Considerations have to be made for dealing with aleatory uncertainty related to the knowledge of the soil conditions. It is shown in the paper, a careful comparison and analysis of the results obtained by different methodologies provides the basis of the selection of practicable methods for the safety analysis of nuclear power plant for beyond design basis liquefaction hazard.

  7. 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.

  8. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 1 - guideword applicability and method description.

    Science.gov (United States)

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

    2008-12-15

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

  9. 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.

  10. DESIGN PACKAGE 1D SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    L.R. Eisler

    1995-02-02

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

  11. Safety analysis for complex systems

    Science.gov (United States)

    Onesty, J. P.; Peercy, R. L., Jr.

    1981-01-01

    Operational risk assessment considers hardware, environment, and human factors. Technique starts with division of postulated mission into segments which are further subdivided into separate operational steps. Consequences of steps, nonoccurrence, premature operation, out-of-sequence operation, and inadvertent execution are examined at subevent, event, and phase levels. Hazards are identified and treated individually. Analysis is well suited to application in energy and transportation fields.

  12. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

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

  13. 14 CFR 33.75 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... direction to that commanded by the pilot; (iv) Uncontrolled fire; (v) Failure of the engine mount system... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.75 Safety analysis. (a) (1) The applicant must analyze the engine, including the control system, to assess the...

  14. 14 CFR 35.15 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: PROPELLERS Design and Construction § 35.15 Safety analysis. (a)(1) The applicant must analyze the propeller system to assess the likely consequences of all failures that can reasonably be expected to...

  15. Evaluation of a non-targeted "Omic"' approach in the safety assessment of genetically modified plants

    DEFF Research Database (Denmark)

    Metzdorff, Stine Broeng; Kok, E. J.; Knuthsen, Pia;

    2006-01-01

    Genetically modified plants must be approved before release in the European Union, and the approval is generally based upon a comparison of various characteristics between the transgenic plant and a conventional counterpart. As a case study, focusing on safety assessment of genetically modified...... plants, we here report the development and characterisation of six independently transformed Arabidopsis thaliana lines modified in the flavonoid biosynthesis. Analyses of integration events and comparative analysis for characterisation of the intended effects were performed by PCR, quantitative Real......, no unintended effects were identified. However, we found that the majority of genes showing differential expression were identified as stress-related genes and that environmental conditions had a large impact on the expression of several genes, proteins, and metabolites. We suggest that the microarray approach...

  16. People detection in nuclear plants by video processing for safety purpose

    Energy Technology Data Exchange (ETDEWEB)

    Jorge, Carlos Alexandre F.; Mol, Antonio Carlos A., E-mail: calexandre@ien.gov.b, E-mail: mol@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN), Rio de Janeiro, RJ (Brazil); Seixas, Jose M.; Silva, Eduardo Antonio B., E-mail: seixas@lps.ufrj.b, E-mail: eduardo@lps.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Eletrica; Cota, Raphael E.; Ramos, Bruno L., E-mail: brunolange@poli.ufrj.b [Universidade Federal do Rio de Janeiro (EP/UFRJ), RJ (Brazil). Dept. de Engenharia Eletronica e de Computacao

    2011-07-01

    This work describes the development of a surveillance system for safety purposes in nuclear plants. The final objective is to track people online in videos, in order to estimate the dose received by personnel, during the execution of working tasks in nuclear plants. The estimation will be based on their tracked positions and on dose rate mapping in a real nuclear plant at Instituto de Engenharia Nuclear, Argonauta nuclear research reactor. Cameras have been installed within Argonauta's room, supplying the data needed. Both video processing and statistical signal processing techniques may be used for detection, segmentation and tracking people in video. This first paper reports people segmentation in video using background subtraction, by two different approaches, namely frame differences, and blind signal separation based on the independent component analysis method. Results are commented, along with perspectives for further work. (author)

  17. Steam generator tube degradation at the Doel 4 plant influence on plant operation and safety

    Energy Technology Data Exchange (ETDEWEB)

    Scheveneels, G. [AIB-Vincotte Nuclear, Brussels (Belgium)

    1997-02-01

    The steam generator tubes of Doel 4 are affected by a multitude of corrosion phenomena. Some of them have been very difficult to manage because of their extremely fast evolution, non linear evolution behavior or difficult detectability and/or measurability. The exceptional corrosion behavior of the steam generator tubes has had its drawbacks on plant operation and safety. Extensive inspection and repair campaigns have been necessary and have largely increased outage times and radiation exposure to personnel. Although considerable effort was invested by the utility to control corrosion problems, non anticipated phenomena and/or evolution have jeopardized plant safety. The extensive plugging and repairs performed on the steam generators have necessitated continual review of the design basis safety studies and the adaptation of the protection system setpoints. The large asymmetric plugging has further complicated these reviews. During the years many preventive and recently also defence measures have been implemented by the utility to manage corrosion and to decrease the probability and consequences of single or multiple tube rupture. The present state of the Doel 4 steam generators remains troublesome and further examinations are performed to evaluate if continued operation until June `96, when the steam generators will be replaced, is justified.

  18. Views on safety culture at Swedish and Finnish nuclear power plants; Syn paa saekerhetskultur vid svenska och finska kaernkraftverk

    Energy Technology Data Exchange (ETDEWEB)

    Hammar, L. [ES-konsulent, (Sweden); Wahlstroem, B.; Kettunen, J. [VTT Automation (Finland)

    2000-02-01

    The report presents the results of interviews about safety culture at Swedish and Finnish nuclear power plants. The aim is to promote the safety work and increase the debate about safety in nuclear power plants, by showing that the safety culture is an important safety factor. The interviews point out different threats, which may become real. It is therefor necessary that the safety aspects get support from of the society and the power plant owners. (EHS)

  19. 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.

  20. 77 FR 15399 - Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task Force...

    Science.gov (United States)

    2012-03-15

    ... COMMISSION Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task Force... Regulatory Commission (NRC) is announcing the availability of the model safety evaluation (SE) for plant..., Revision 1, is available in ADAMS under Accession No. ML111650552; the model application is available...

  1. 77 FR 27814 - Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task Force...

    Science.gov (United States)

    2012-05-11

    ... COMMISSION Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task Force... availability. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is announcing the availability of the model safety evaluation (SE) for plant-specific adoption of Technical Specifications (TSs) Task Force...

  2. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant,...

  3. 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)

  4. Safety Culture Assessment in Petrochemical Industry: A Comparative Study of Two Algerian Plants

    Directory of Open Access Journals (Sweden)

    Assia Boughaba

    2014-06-01

    Conclusion: The comparison between the two petrochemical plants of the group Sonatrach confirms these results in which Company A, the managers of which are English and Norwegian, distinguishes itself by the maturity of their safety culture has significantly higher evaluations than the company B, who is constituted of Algerian staff, in terms of safety management practices and safety performance.

  5. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    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.

  6. Perspective of regulation on software safety analysis: experience of software safety analysis activity of Lungmen project

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chuan Chung [Taiwan Power Company, Taipei TW (China)

    2005-11-15

    Software Safety Analysis is one of the essential tasks must be performed in the design work of digital computer software used in safety system of Nuclear Power Station. While there is more experience in Software Verification and Validation and Configuration Management in software industry, Software Safety Analysis (SSA) is a new task. What is the scope of SSA? What should be done in SSA? Various SSA related code and Standards were reviewed and from the evolvement of code and standards, it was concluded that Abnormal Condition and Events should be treated as part of SSA activities and SSA activities could be one of the activities in Software V and V SSA case study on NUMAC as Pervious Developed System was presented and a new method on SSA - 'Hazard Analysis and Defense in Depth for Software Safety Analysis' to enhance the confidence in SSA activities in Lungmen project was introduced.

  7. Preliminary report: in-plant safety/relief valve discharge load test, Monticello Plant

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H.C. (comp.)

    1976-12-01

    This preliminary report covers the results of the test program of safety/relief valve (SRV) discharge load phenomena and the effects upon the Mark I primary containment torus structure of the Monticello Nuclear Power Plant. The objectives of the test were to provide a data base for verifying/improving analytical models and to measure the structural response of the torus to SRV discharges. Objectives, instrumentation, and test plan are described. Results of continuing data evaluation will be included in the final report scheduled for publication later in 1977.

  8. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-05-01

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

  9. Determining safety criteria for reinforced concrete structures of power plants taking the example of a nuclear power plant with RBMK

    Directory of Open Access Journals (Sweden)

    Nikolayev Valery

    2016-01-01

    Full Text Available The paper shows how safety criteria of nuclear power plants with reactor RBMK can be defined based on analytical, numerical and mixed calculation methods using data about strength characteristics of materials with the course of time.

  10. 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.

  11. Station Blackout: A case study in the interaction of mechanistic and probabilistic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Diego Mandelli; Cristian Rabiti

    2013-11-01

    The ability to better characterize and quantify safety margins is important to improved decision making about nuclear power plant design, operation, and plant life extension. As research and development (R&D) in the light-water reactor (LWR) Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway R&D is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario.

  12. 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.

  13. Safety of GM crops: compositional analysis.

    Science.gov (United States)

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

    2013-09-04

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

  14. Safety Assessment of Low-Contaminated Equipment Dismantling at Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Egidijus Babilas

    2015-01-01

    Full Text Available The decommissioning of nuclear facilities requires adequate planning and demonstration that dismantling and decontamination activities can be conducted safely. Existing safety standards require that an appropriate safety assessment be performed to support the decommissioning plan for each facility (International Atomic Energy Agency, 2006. This paper presents safety assessment approach used in Lithuania during the development of the first dismantling and decontamination project for Ignalina NPP. The paper will mainly focus on the identification and assessment of the hazards raised due to dismantling and decontamination activities at Ignalina Nuclear Power Plant and on the assessment of the nonradiological and radiological consequences of the indicated most dangerous initiating event. The drop of heavy item was indicated as one of most dangerous initiating events for the discussed Ignalina Nuclear Power Plant dismantling and decontamination project. For the analysis of the nonradiological impact the finite element model for the load drop force calculation was developed. The radiological impact was evaluated in those accident cases which would lead to the worst radiological consequences. The assessments results show that structural integrity of the building and supporting columns of building structures will be maintained and radiological consequences are lower than the annual regulatory operator dose limit.

  15. [Health & safety in a steel plant: technical and managerial action].

    Science.gov (United States)

    Fusato, M

    2012-01-01

    The report presents the experience in a steel company to improve the management of issues relating to health and safety of workers. The first part of the work focuses on the description of the interventions made by the company in recent years, which can be divided into technical interventions on production facilities, training and information, organizational activities and specific projects in collaboration with the health service. The second part presents the certification project according to OHSAS 18001, with particular focus on the efforts for a lean management of the documentation required by the management systems and for the automation of internal processes. The last part finally describes in detail the manner in which it was decided to address some issues that significantly affect the factory doctor: the recording and analysis of accidents and medications, management of hazardous substances and personal protective equipment.

  16. Chirospecific analysis of plant volatiles

    Energy Technology Data Exchange (ETDEWEB)

    Tkachev, A V [N.N. Vorozhtsov Novosibirsk Institute of Organic Chemistry, Siberian Branch of the Russian Academy of Sciences, Novosibirsk (Russian Federation)

    2007-10-31

    Characteristic features of the analysis of plant volatiles by enantioselective gas (gas-liquid) chromatography and gas chromatography/mass spectrometry are discussed. The most recent advances in the design of enantioselective stationary phases are surveyed. Examples of the preparation of the most efficient phases based on modified cyclodextrins are given. Current knowledge on the successful analytical resolution of different types of plant volatiles (aliphatic and aromatic compounds and mono-, sesqui- and diterpene derivatives) into optical antipodes is systematically described. Chiral stationary phases used for these purposes, temperature conditions and enantiomer separation factors are summarised. Examples of the enantiomeric resolution of fragrance compounds and components of plant extracts, wines and essential oils are given.

  17. Chirospecific analysis of plant volatiles

    Science.gov (United States)

    Tkachev, A. V.

    2007-10-01

    Characteristic features of the analysis of plant volatiles by enantioselective gas (gas-liquid) chromatography and gas chromatography/mass spectrometry are discussed. The most recent advances in the design of enantioselective stationary phases are surveyed. Examples of the preparation of the most efficient phases based on modified cyclodextrins are given. Current knowledge on the successful analytical resolution of different types of plant volatiles (aliphatic and aromatic compounds and mono-, sesqui- and diterpene derivatives) into optical antipodes is systematically described. Chiral stationary phases used for these purposes, temperature conditions and enantiomer separation factors are summarised. Examples of the enantiomeric resolution of fragrance compounds and components of plant extracts, wines and essential oils are given.

  18. 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

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

    National Research Council Canada - National Science Library

    Zongzhi Li; Hoang Dao; Harshingar Patel; Yi Liu; Bei Zhou

    2017-01-01

    .... This study introduces a refined method for computing the Safety Index (SI) as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis...

  20. Evolution of safety systems in the Juzbado Plant; Evolucion de los sistemas de seguridad en la fabrica de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    Merino, M. L.

    2010-07-01

    Safety has always been one of the cornerstones of the operation of the Juzbado Plant since it began operations in 1985. Along these 25 years the systems for monitoring and controlling the impact of the Plants operation have evolved significantly, always trying to maintain the highest standards of the nuclear industry in all the activities with relevance to safety. The drivers of these developments were changes in regulations, the increase of the production capacity of the Plant, the design modifications incorporated in its facilities and of course, the lessons learned during the years of operation. Furthermore, over the last few years the Juzbado Plant has developed and implemented several analysis methodologies to cope with specific issues regarding safety management. The latest is programme for the Systematic Review of the Operational Conditions of the Safety Systems, which was initiated in 2009 to assess the actual operating conditions of all the system with safety relevance, aiming to identify areas for improvement in order to ensure their high-performance after 25 years of operation. Following more significant examples of this evolution are showing. (Author)

  1. 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.

  2. DEVELOPMENT OF METHODS IMPROVING INDUSTRIAL SAFETY OF TECHNOLOGICAL PROCESSES IN ASPHALT-CONCRETE PLANT MIXERS

    Directory of Open Access Journals (Sweden)

    I. A. Ivanova

    2010-05-01

    Full Text Available Problem statement. The problem of improvement of industrial safety of technol-ogical processes in mixers of asphalt-concrete plants is considered on the basis of analysis of organic impurities content in incomplete combustion products, and es-timation of efficiency of purification of asphalt-concrete plant emissions in the presence of “wet” flue gas purification system is given.Results and conclusions. It has been found that the efficiency of hydrocarbon fuel burning affects the amount of hydrophobic dust thrown into the atmosphere, and burning of heavy fuel oil is attended by significant incompleteness of fuel combustion, and this is connected with the processes of fuel dispersion and evapo-ration. The optimal measures for efficient combustion and cleaning of hydrophob-ic dust are described.

  3. 福岛核事故后核电厂安全改进行动分析%Analysis of Nuclear Power Plant Safety Improvement Action After Fukushima Daiichi NPP Accident

    Institute of Scientific and Technical Information of China (English)

    张琳; 李文宏; 杨红义

    2014-01-01

    介绍了福岛核事故后世界上主要核电国家相继开展的核电厂安全检查、再评价行动,并得出相应的检查和测试结论。法国、美国和中国等国家分别提出了福岛核事故后改进核电厂安全的建议、要求和行动,并制定了具体工程措施:在极端外部事件的设防,严重事故预防和缓解,水、电、通风实体改进,限制严重事故下的放射性释放和应急准备等主要方面开展的安全改进行动,将会提高核电厂的安全水平并提升缓解严重事故的能力。反思福岛核事故,总结福岛核事故对核电安全技术改进的促进作用,对未来核电安全技术的发展进行了展望。%The paper introduces nuclear power plant (NPP) safety inspections ,actions of re-evaluation and the main conclusions of the main countries with nuclear power in the world after the Fukushima Daiichi NPP accident .Countries ,such as France ,the United States and China ,carried out the NPP safety inspection and re-evaluation action ,and acquired conclusions . These countries respectively put forward the suggestions , demands and actions about improving NPP safety , and also formulated the specific engineering measures .The safety improvement actions ,such as fortification in extreme external events , severe accident prevention and mitigation , water , electricity and ventilation physical improvements ,limiting the radioactive release of severe accident , emergency preparedness ,and so on ,will increase the safety level and enhance ability to alleviate the severe accident of NPP .The profound consideration about the Fukushima Daiichi NPP accident , summarizing the promotion role of nuclear power safety technology improvement after the Fukushima Daiichi NPP accident , and the development of nuclear power safety technology in the future were discussed .

  4. Treatment of the safety culture in the Integrated Plant Supervision System (SISC); Tratamiento de la cultura de seguridad en el Sistema Integrado de Supervision de Centrales (SISC)

    Energy Technology Data Exchange (ETDEWEB)

    Barrientos Montero, M.; Gil Ontes, B.

    2013-09-01

    Certain of the catastrophes that occurred in the 1980's, among them the Chernobyl accident, underlined the importance of the human factor and led to the development of the concept of the safety culture. Today this is one of the main aspects involved in safety analysis, as applied in the management of nuclear power plants. (Author)

  5. Treatment of the safety culture in the Integrated Plant Supervision System; Tratamiento de la cultura de seguridad en el Sistema Integrado de Supervision de Centrales (SISC)

    Energy Technology Data Exchange (ETDEWEB)

    Barrientos Montero, M.; Gil Montes, B.

    2013-06-01

    Certain of the catastrophes that occurred in the 1980's, among them the Chernobyl accident, underlined the importance of the human factor and led to the development of the concept of the safety culture. Today this is one of the main aspects involved in safety analysis, as applied in the management of nuclear power plants. (Author)

  6. Evaluation on safety concerns of integral reactor: development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Lee, S. G.; Sin, A. D. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They includes the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. These efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. 62 refs., 3 figs., 21 tabs. (Author)

  7. SAFIR2014. The Finnish Research Programme on Nuclear Power Plant Safety 2011-2014. Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. (ed.)

    2013-02-15

    The Finnish Nuclear Power Plant Safety Research Programme 2011-2014, SAFIR2014, is a 4-year publicly funded national technical and scientific research programme on the safety of nuclear power plants. The programme is funded by the State Nuclear Waste Management Fund (VYR), as well as other key organisations operating in the area of nuclear energy. The programme provides the necessary conditions for retaining knowledge needed for ensuring the continuance of safe use of nuclear power, for developing new know-how and for participation in international co-operation. The SAFIR2014 Steering Group, responsible of the strategic alignements of the programme, consists of representatives of the Finnish Nuclear Safety Authority (STUK), Ministry of Employment and the Economy (MEE), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oyj (TVO), Fortum Power and Heat Oy (Fortum), Fennovoima Oy, Lappeenranta University of Technology (LUT), Aalto University (Aalto), Finnish Funding Agency for Technology and Innovation (Tekes), Finnish Institute of Occupational Health (TTL) and the Swedish Radiation Safety Authority (SSM). The research programme is divided into nine areas: Man, organisation and society, Automation and control room, Fuel research and reactor analysis, Thermal hydraulics, Severe accidents, Structural safety of reactor circuits, Construction safety, Probabilistic risk analysis (PRA), and Development of research infrastructure. A reference group is assigned to each of these areas to respond for the strategic planning and to supervise the projects in its respective field. Research projects are selected annually based on a public call for proposals. Most of the projects are planned for the entire duration of the programme, but there can also be shorter one- or two-year projects. The annual volume of the SAFIR2014 programme in 2011-2012 has been 9,5-9,9 M euro. Main funding organisations were the State Nuclear Waste Management Fund (VYR) with over 5 M euro and

  8. Safety Management Analysis In Construction Industry

    Directory of Open Access Journals (Sweden)

    T. Subramani

    2014-06-01

    Full Text Available 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, safety incentives and penalties, workers’ attitude towards safety, labor turnover rates and compliance with safety legislation. The study will also reveal several factors of poor safety management. Thus the paper will conclude by providing a set of recommendations and strategies to contractors for improving their safety performance.

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis L; Mandelli, Diego; Zhegang Ma

    2014-11-01

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

  10. A cross-cultural study of organizational factors on safety: Japanese vs. Taiwanese oil refinery plants.

    Science.gov (United States)

    Hsu, Shang Hwa; Lee, Chun-Chia; Wu, Muh-Cherng; Takano, Kenichi

    2008-01-01

    This study attempts to identify idiosyncrasies of organizational factors on safety and their influence mechanisms in Taiwan and Japan. Data were collected from employees of Taiwanese and Japanese oil refinery plants. Results show that organizational factors on safety differ in the two countries. Organizational characteristics in Taiwanese plants are highlighted as: higher level of management commitment to safety, harmonious interpersonal relationship, more emphasis on safety activities, higher devotion to supervision, and higher safety self-efficacy, as well as high quality of safety performance. Organizational characteristics in Japanese plants are highlighted as: higher level of employee empowerment and attitude towards continuous improvement, more emphasis on systematic safety management approach, efficient reporting system and teamwork, and high quality of safety performance. The casual relationships between organizational factors and workers' safety performance were investigated using structural equation modeling (SEM). Results indicate that the influence mechanisms of organizational factors in Taiwan and Japan are different. These findings provide insights into areas of safety improvement in emerging countries and developed countries respectively.

  11. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

  12. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Kwon, Y. M.; Suk, S. D. [and others

    2005-03-01

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

  13. 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.

  14. Implementation of Recommendations from the One System Comparative Evaluation of the Hanford Tank Farms and Waste Treatment Plant Safety Bases

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, Richard L.; Niemi, Belinda J.; Paik, Ingle K.; Buczek, Jeffrey A.; Lietzow, J.; McCoy, F.; Beranek, F.; Gupta, M.

    2013-11-07

    A Comparative Evaluation was conducted for One System Integrated Project Team to compare the safety bases for the Hanford Waste Treatment and Immobilization Plant Project (WTP) and Tank Operations Contract (TOC) (i.e., Tank Farms) by an Expert Review Team. The evaluation had an overarching purpose to facilitate effective integration between WTP and TOC safety bases. It was to provide One System management with an objective evaluation of identified differences in safety basis process requirements, guidance, direction, procedures, and products (including safety controls, key safety basis inputs and assumptions, and consequence calculation methodologies) between WTP and TOC. The evaluation identified 25 recommendations (Opportunities for Integration). The resolution of these recommendations resulted in 16 implementation plans. The completion of these implementation plans will help ensure consistent safety bases for WTP and TOC along with consistent safety basis processes. procedures, and analyses. and should increase the likelihood of a successful startup of the WTP. This early integration will result in long-term cost savings and significant operational improvements. In addition, the implementation plans lead to the development of eight new safety analysis methodologies that can be used at other U.S. Department of Energy (US DOE) complex sites where URS Corporation is involved.

  15. 77 FR 58421 - Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task Force...

    Science.gov (United States)

    2012-09-20

    ... COMMISSION Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task Force...-415- 4737, or by email to pdr.resource@nrc.gov . TSTF-522, Revision 0, includes a model application and is available in ADAMS under Accession No. ML100890316. The model safety evaluation (SE) of...

  16. The application of transcriptomics in the comparative safety assessment of (GMO-derived) plant products

    NARCIS (Netherlands)

    Kok, E.J.

    2008-01-01

    National and international organizations have discussed current approaches to the safety assessment of complex (plant) food products in general and the safety assessment of GMO-derived food products in particular. One of the recommendations of different expert meetings was that the new analytical te

  17. The Probabilistic Safety Analysis during low power and shutdown, framework to improve safety; El APS a baja potencia en parada, marco para la mejora de la seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Nos, V.

    2014-02-01

    Historically Probabilistic Safety Analysis (PSA) has been focused exclusively at full power operation, nevertheless, operational experience has revealed that events occurred during low power and shutdown can also present threats for the safety of the plant. Through qualitative assessment (NUMARC 91-06) about the configuration in shutdown have been internationally accepted, the benefits of Low Power and Shutdown PSA have been demonstrated as fundamental framework of quantitative understanding for improving safety and risk management in the above mentioned operative conditions of the plant. (Author)

  18. 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)

  19. Investigation of the performance based structural safety factor of elbow pipes in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Ho; Park, Chi Yong [Korea Electric Power Reserch Institute, Daejeon (Korea, Republic of); Park, Jai Hak [Chungbuk National University, Cheongju (Korea, Republic of)

    2009-07-01

    The piping systems in nuclear power plant are composed of various typed pipes such as straight pipe, elbow, branch and reducer etc. The elbow is connected from straight pipe to another pipes in order to establish the complicated piping system. Elbow is one of very important components considering management of wall thinning degradation. It is however applied by various loads such as system pressure, earthquake, postulated break loading and many transient loads, which provoke simply the internal pressure, bending and torsional stress. In this study, firstly pipes in the secondary system of the nuclear power plant are investigated in view of the ratio of radius to thickness. Next, a large number of finite element analysis considering the all typed dimensions of commercial pipe has been performed to find out the behavior of TES(Twice Elastic Slope) plastic load of elbows, which is based on evaluation of the structural safety factor. Finally performance based structural safety factor was investigated comparing with maximum allowable load by construction code.

  20. 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.

  1. 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.

  2. Nuclear power plants and safety; Elektrownie jadrowe i bezpieczenstwo

    Energy Technology Data Exchange (ETDEWEB)

    Celinski, Z. [Politechnika Warszawska, Warsaw (Poland)

    1995-12-31

    The brief scope on the state of nuclear energetics worldwide as well as development perspectives have been presented. The safety problems, economic competitiveness and public acceptance have been shown and discussed. 55 refs, 3 figs, 2 tabs.

  3. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-03-01

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

  6. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

  7. Licensee responsibility for nuclear power plant safety; Verantwortung der Genehmigungsinhaber fuer die Sicherheit der Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Horst

    2010-02-15

    Simple sentences easy to grasp are desirable in regulations and bans. However, in a legal system, their meaning must be unambiguous. Article 6, Paragraph 1 of the EURATOM Directive on a community framework for the nuclear safety of nuclear facilities of June 2009 states that 'responsibility for the nuclear safety of a nuclear facility is incumbent primarily on the licensee.' The draft 'Safety Criteria for Nuclear Power Plants, Revision D, April 2009' of the German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) (A Module 1, 'Safety Criteria for Nuclear Power Plants: Basic Safety Criteria' / '0 Principles' Paragraph 2) reads: 'Responsibility for ensuring safety rests with the licensee. He shall give priority to compliance with the safety goal over the achievement of other operational objectives.' In addition, the existing rules and regulations, whose rank is equivalent to that of international regulations, assign priority to the safety goal to be pursued by the licensee over all other objectives of the company. The operator's responsibility for nuclear safety can be required and achieved only on the basis of permits granted, which must meet legal requirements. The operator's proximity to plant operation is the reason for his 'primary responsibility.' Consequently, verbatim incorporation of Article 6, Paragraph 1 of the EURATOM Directive would only be a superscript added to existing obligations of the operator - inclusive of a safety culture designed as an incentive to further 'the spirit of safety-related actions' - without any new legal contents and consequences. In the reasons of the regulation, this would have to be clarified in addition to the cryptic wording of 'responsibility.. primarily,' at the same time expressing that operators and authorities work together in a spirit of openness and trust. (orig.)

  8. Phytochemical analysis of selected medicinal plants | Hussain ...

    African Journals Online (AJOL)

    Phytochemical analysis of selected medicinal plants. ... African Journal of Biotechnology ... Abstract. Four medicinal plants including Ranunculus arvensis, Equisetum ravens, Carathamus lanatus and Fagonia critica were used for the study.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy (Finland); Bjoere, S.; Olsson, Lena [ABB Atom AB, Vaesteraas (Sweden)

    1992-12-01

    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.

  10. Living PSA program: LIPSAS development for safety management of an LMFBR plant

    Energy Technology Data Exchange (ETDEWEB)

    Aizawa, Kiyoto [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan); Nakai, Ryodai [O-arai Engineering Center, Ibaraki (Japan)

    1994-12-31

    During construction and subsequent operation of a nuclear power plant, many changes occur in components, systems and operating procedures, which continuously modify the configuration of the power plant. A living PSA program can assess and manage safety-related operations and plant changes by adequately reproducing plant models and structured databases corresponding to the changes in system configuration. A living PSA system, LIPSAS, has been developed for the Japanese prototype liquid metal-cooled fast-breeder reactor (LMFBR), Monju, which is in the preoperation functional test stage. In order to utilize the LIPSAS as a risk management tool, equations for the schematic time history of the plant risk level and the relative risk criteria have been developed. Experience with LIPSAS shows that this program is a prospective tool to support decisions that affect plant safety, although a continuing and significant resource commitment of the operations staff at the site is still required. (author).

  11. Japan`s international cooperation programs on seismic safety of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Sanada, Akira [Agency of Natural Resources and Energy, Tokyo (Japan)

    1997-03-01

    MITI is promoting many international cooperation programs on nuclear safety area. The seismic safety of nuclear power plants (NPPs) is a one of most important cooperation areas. Experts from MITI and related organization join the multilateral cooperation programs carried out by international organization such as IAEA, OECD/NEA etc. MITI is also promoting bilateral cooperation programs such as information exchange meetings, training programs and seminars on nuclear safety with several countries. Concerning to the cooperation programs on seismic safety of NPPs such as information exchange and training, MITI shall continue and expand these programs. (J.P.N.)

  12. Safety requirements to the operation of hydropower plants; Sicherheit beim Betrieb von Wasserkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Lux, Reinhard [Berufsgenossenschaft Energie Textil Elektro Medienerzeugnisse (BG ETEM), Koeln (Germany)

    2011-07-01

    Employers have to take into account various safety and health requirements relating to the design, construction, operation and maintenance of hydropower plants. Especially the diversity of the hydropower plant components requires the consideration of different safety and health aspects. In 2011 the ''Fachausschuss Elektrotechnik'' (expert committee electro-technics) of the institution for statutory accident insurance and prevention presented a new ''BG-Information'' dealing with ''Safe methods operating hydropower plants''. The following article gives an introduction into the conception and the essential requirements of this new BG-Information. (orig.)

  13. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

  14. Safety culture in nuclear power plants. Proceedings; Sicherheitskultur im Kernkraftwerk. Seminarbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-01

    As a consequence of the INSAG-4 report on `safety culture`, published by the IAEA in 1991, the Federal Commission for the Safety of Nuclear Power Plants (KSA) decided to hold a one-day seminar as a first step in this field. The KSA is an advisory body of the Federal Government and the Federal Department of Transport and Energy (EVED). It comments on applications for licenses, observes the operation of nuclear power plants, assists with the preparation of regulations, monitors the progress of research in the field of nuclear safety, and makes proposals for research tasks. The objective of this seminar was to familiarise the participants with the principles of `safety culture`, with the experiences made in Switzerland and abroad with existing concepts, as well as to eliminate existing prejudices. The main points dealt with at this seminar were: - safety culture from the point of view of operators, - safety culture from the point of view of the authorities, - safety culture: collaboration between power plants, the authorities and research organisations, - trends and developments in the field of safety culture. Invitations to attend this seminar were extended to the management boards of companies operating Swiss nuclear power plants, and to representatives of the Swiss authorities responsible for the safety of nuclear power plants. All these organisations were represented by a large number of executive and specialist staff. We would like to express our sincerest thanks to the Head of the Federal Department of Transport and Energy for his kind patronage of this seminar. (author) figs., tabs., refs.

  15. Atomic Information Technology Safety and Economy of Nuclear Power Plants

    CERN Document Server

    Woo, Taeho

    2012-01-01

    Atomic Information Technology revaluates current conceptions of the information technology aspects of the nuclear industry. Economic and safety research in the nuclear energy sector are explored, considering statistical methods which incorporate Monte-Carlo simulations for practical applications. Divided into three sections, Atomic Information Technology covers: • Atomic economics and management, • Atomic safety and reliability, and • Atomic safeguarding and security. Either as a standalone volume or as a companion to conventional nuclear safety and reliability books, Atomic Information Technology acts as a concise and thorough reference on statistical assessment technology in the nuclear industry. Students and industry professionals alike will find this a key tool in expanding and updating their understanding of this industry and the applications of information technology within it.

  16. Food plant toxicants and safety - Risk assessment and regulation of inherent toxicants in plant foods

    DEFF Research Database (Denmark)

    Essers, A.J.A.; Alink, G.M.; Speijers, G.J.A.

    1998-01-01

    The ADI as a tool for risk management and regulation of food additives and pesticide residues is not readily applicable to inherent food plant toxicants: The margin between actual intake and potentially toxic levels is often small; application of the default uncertainty factors used to derive ADI...... values, particularly when extrapolating from animal data, would prohibit the utilisation of the food, which may have an overall beneficial health effect. Levels of inherent toxicants are difficult to control; their complete removal is not always wanted, due to their function for the plant or for human...... health. The health impact of the inherent toxicant is often modified by factors in the food, e.g. the bioavailability from the matrix and interaction with other inherent constituents. Risk-benefit analysis should be made for different consumption scenarios, without the use of uncertainty factors. Crucial...

  17. Analysis of road safety management systems in Europe.

    NARCIS (Netherlands)

    Muhlrad, N. Vallet, G. Butler, I. Gitelman, V. Doveh, E. Dupont, E. Thomas, P. Talbot, R. Papadimitriou, E. Yannis, G. Persia, L. Giustiniani, G. Machata, K. & Bax, C.A.

    2014-01-01

    The objective of this paper is the analysis of road safety management in European countries and the identification of “good practice”. A road safety management investigation model was created, based on several “good practice” criteria. Road safety management systems have been thoroughly investigated

  18. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  19. 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.

  20. The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Guerpinar, A.; Godoy, A. [International Atomic Energy Agency, Vienna (IAEA). Div. of Nuclear Installation Safety

    1997-03-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on `Benchmark study for the seismic analysis and testing of WWER type nuclear power plants`. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

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

    Directory of Open Access Journals (Sweden)

    Zongzhi Li

    2017-04-01

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

  2. SAFIR2010. The Finnish research programme on nuclear power plant safety 2007-2010. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. (ed.)

    2009-02-15

    Major part of Finnish public research on nuclear power plant safety during the years 2007-2008 has been carried out in the SAFIR2010 programme. The steering group of SAFIR2010 consists of representatives from Radiation and Nuclear Safety Authority (STUK), Ministry of Employment and the Economy (MEE), VTT Technical Research Centre of Finland (VTT), Teollisuuden Voima Oyj (TVO), Fortum Power and Heat Oyj, Fortum Nuclear Services Oy (Fortum), Tekes - the Finnish Funding Agency for Technology and Innovation (Tekes), Helsinki University of Technology (TKK) and Lappeenranta University of Technology (LUT). In addition to representatives of these organisations, the Steering Group has permanent experts from the Swedish Radiation Safety Authority (SSM) and Fennovoima Oy (Fennovoima). SAFIR2010 research programme is divided in eight research areas that are Organisation and human, Automation and control room, Fuel and reactor physics, Thermal hydraulics, Severe accidents, Structural safety of reactor circuit, Construction safety, and Probabilistic Safety Analysis (PSA). Research projects of the programme are chosen on the basis of annual call for proposals. The annual volume of the SAFIR2010 programme in 2007-2008 has been 6,3-6,7 M euro and approximately 50 person years. Main funding organisations in 2007-2008 were State Waste Management Fund VYR with 2,7-3,0 M euro and VTT with 2,4-2,5 M euro annually. In 2008 research was carried out in 30 projects. The research in the programme has been carried out primarily by VTT Technical Research Centre of Finland. Other research units responsible for the projects solely or in co-operation with other institutions include Lappeenranta University of Technology, Helsinki University of Technology, Tampere University of Technology, Fortum Nuclear Services Oy, Finnish Institute of Occupational Health and Finnish Meteorological Institute. In addition, there have been a few minor subcontractors in some projects. The programme management

  3. SAFIR2010. The Finnish Research Programme on Nuclear Power Plant Safety 2007-2010. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K.; Suolanen, V. (eds.)

    2011-02-15

    Major part of Finnish public research on nuclear power plant safety during the years 2007-2010 has been carried out in the SAFIR2010 programme. The steering group of SAFIR2010 consisted of representatives from Radiation and Nuclear Safety Authority (STUK), Ministry of Employment and the Economy (MEE), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oyj (TVO), Fortum Power and Heat Oyj, Fortum Nuclear Services Oy (Fortum), Finnish Funding Agency for Technology and Innovation (Tekes), Aalto University School of Science and Technology (Aalto, former Helsinki University of Technology) and Lappeenranta University of Technology (LUT). In addition to representatives of these organisations, the Steering Group had permanent experts from the Swedish Radiation Safety Authority (SSM) and Fennovoima Oy (Fennovoima). SAFIR2010 research programme was divided in eight research areas that were Organisation and human, Automation and control room, Fuel and reactor physics, Thermal hydraulics, Severe accidents, Structural safety of reactor circuit, Construction safety, and Probabilistic Safety Analysis (PSA). Research projects of the programme were chosen on the basis of annual call for proposals. The annual volume of the SAFIR2010-programme in 2007-2010 has been 6,5-7,1 M euro and approximately 50 person years. Main funding organisations in 2007-2010 have been the State Waste Management Fund VYR with 2,7-3,0 M euro and VTT with 2,4-2,7 M euro annually. In 2010 research was carried out in 33 projects. The research in the programme has been carried out primarily by VTT Technical Research Centre of Finland. Other research units responsible for the projects solely or in co-operation with other institutions include Lappeenranta University of Technology, Aalto University (previously Helsinki University of Technology), Tampere University of Technology, Fortum Power and Heat Oy (previously Fortum Nuclear Services Oy), Finnish Institute of Occupational Health and Finnish

  4. Assessment of effects of fires on safety of nuclear power plants. Paloturvallisuuden arviointi ydinvoimalassa

    Energy Technology Data Exchange (ETDEWEB)

    Keski-Rahkonen, O.

    1992-01-01

    Experience and probabilistic safety assessments have shown that fires may present a major hazard in a nuclear plant either as initial events or as a factor aggravating the consequences from accidents initiated otherwise. Numerical modelling of fires can be performed in various ways. The oldest approach is based on experimental models where rough correlations are employed. Depending on the type of application more advanced codes are employed in fire analyses. In zone models each compartment is divided into two horizontal layers, which both are at the same temperature. In system models the building to be analyzed is divided into interconnected nodes. The most complicated fire analysis models are field models, which calculate multidimensional fields of temperatures and other quantities by solving numerically the conservation equations for several variables.

  5. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  6. 10 CFR 52.157 - Contents of applications; technical information in final safety analysis report.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Contents of applications; technical information in final safety analysis report. 52.157 Section 52.157 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS Manufacturing Licenses § 52.157 Contents of applications...

  7. 10 CFR 52.79 - Contents of applications; technical information in final safety analysis report.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Contents of applications; technical information in final safety analysis report. 52.79 Section 52.79 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS Combined Licenses § 52.79 Contents of applications...

  8. 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.

  9. Analysis on safety production in coal mines Henan Province

    Institute of Scientific and Technical Information of China (English)

    KONG Liu-an; ZHANG Wen-yong

    2006-01-01

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

  10. Safety features in nuclear power plants to eliminate the need of emergency planning in public domain

    Indian Academy of Sciences (India)

    P K Vijayan; M T Kamble; A K Nayak; K K Vaze; R K Sinha

    2013-10-01

    Following the Fukushima accident, the safety features of Nuclear Power Plants (NPP) are being re-examined worldwide including India to demonstrate capabilities to cope with severe accidents. In order to restore public confidence and support for nuclear power, it is felt necessary to design future NPPs with near zero impact outside the plant boundary and thus enabling elimination of emergency planning in public domain. Authors have identified a set of safety features which are needed to be incorporated in advanced reactors to achieve this goal. These features enabling prevention, termination, mitigation and containment of radioactivity for beyond design basis accidents arising from extreme natural events are essential for achieving the goal of elimination of emergency planning in public domain. Inherent safety characteristics, passive and engineered safety features to achieve these functions are discussed in this paper. Present trends and future developments in this direction are also described briefly.

  11. POEAS: Automated Plant Phenomic Analysis Using Plant Ontology.

    Science.gov (United States)

    Shameer, Khader; Naika, Mahantesha Bn; Mathew, Oommen K; Sowdhamini, Ramanathan

    2014-01-01

    Biological enrichment analysis using gene ontology (GO) provides a global overview of the functional role of genes or proteins identified from large-scale genomic or proteomic experiments. Phenomic enrichment analysis of gene lists can provide an important layer of information as well as cellular components, molecular functions, and biological processes associated with gene lists. Plant phenomic enrichment analysis will be useful for performing new experiments to better understand plant systems and for the interpretation of gene or proteins identified from high-throughput experiments. Plant ontology (PO) is a compendium of terms to define the diverse phenotypic characteristics of plant species, including plant anatomy, morphology, and development stages. Adoption of this highly useful ontology is limited, when compared to GO, because of the lack of user-friendly tools that enable the use of PO for statistical enrichment analysis. To address this challenge, we introduce Plant Ontology Enrichment Analysis Server (POEAS) in the public domain. POEAS uses a simple list of genes as input data and performs enrichment analysis using Ontologizer 2.0 to provide results in two levels, enrichment results and visualization utilities, to generate ontological graphs that are of publication quality. POEAS also offers interactive options to identify user-defined background population sets, various multiple-testing correction methods, different enrichment calculation methods, and resampling tests to improve statistical significance. The availability of such a tool to perform phenomic enrichment analyses using plant genes as a complementary resource will permit the adoption of PO-based phenomic analysis as part of analytical workflows. POEAS can be accessed using the URL http://caps.ncbs.res.in/poeas.

  12. 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.

  13. An Integrated Approach of Model checking and Temporal Fault Tree for System Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Koh, Kwang Yong; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2009-10-15

    Digitalization of instruments and control systems in nuclear power plants offers the potential to improve plant safety and reliability through features such as increased hardware reliability and stability, and improved failure detection capability. It however makes the systems and their safety analysis more complex. Originally, safety analysis was applied to hardware system components and formal methods mainly to software. For software-controlled or digitalized systems, it is necessary to integrate both. Fault tree analysis (FTA) which has been one of the most widely used safety analysis technique in nuclear industry suffers from several drawbacks as described in. In this work, to resolve the problems, FTA and model checking are integrated to provide formal, automated and qualitative assistance to informal and/or quantitative safety analysis. Our approach proposes to build a formal model of the system together with fault trees. We introduce several temporal gates based on timed computational tree logic (TCTL) to capture absolute time behaviors of the system and to give concrete semantics to fault tree gates to reduce errors during the analysis, and use model checking technique to automate the reasoning process of FTA.

  14. 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....

  15. Regulatory and safety aspects of ageing in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kaufer, B. [OECD/Nuclear Energy Agency, Issy-les-Moulineaux (France). Nuclear Safety Div.

    2002-08-01

    The OECD Nuclear Energy Agency (NEA) is a semi-autonomous body within the OECD established in 1958 with the mandate to promote co-operation among the governments of its participating countries in furthering the development of nuclear power as a safe, environmentally acceptable and economicy energy source. While all of groups have detailed programmes involving important aspects, this paper will focus specifically on the work of Committee on Nuclear Regulatory Activities (CNRA) and the Committee on the Safety of Nuclear Installations (CSNI). (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

    In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

  17. On the Seismic Safety of Nuclear Power Plant Sites in South Korea

    Science.gov (United States)

    Choi, H.; Park, S.; Yang, J.; Shim, T.; Im, C. B.

    2016-12-01

    The Korean Peninsula is located at the far eastern part of Eurasian Plate, and within the intra-plate region several hundred km away from the nearest plate boundary. The earthquakes around the Korean Peninsula show the typical characteristics of intra-plate earthquakes. So to speak, those are low seismicity, relatively smaller magnitude than that of inter-plate earthquakes, and spatially irregular epicenters. There are 24 nuclear power plants (NPPs) in operation, 4 NPPs in completion of construction, and 4 NPPs in preparation of construction in South Korea. Even though the seismicity of the Korean Peninsula is known as relatively low, but because there are more than 30 NPPs within not so large territory, thorough the preparedness of NPPs' safety against earthquakes is required. The earthquake preparedness of NPPs in South Korea is composed of 4 stages: site election, design, construction and operation. Since regulatory codes and standards are strictly applied in each stage, the NPPs in South Korea are believed to be safe enough against the maximum potential earthquake ground motion. Through data analysis on geological and seismological characteristics of the region within a radius of 320 km from the site and the detailed geological survey of the area within a radius of 8 km from the site, the design earthquake ground motion of NPPs in South Korea is determined to be 0.2g (in case of newly constructed NPPs is 0.3g) considering the maximum potential earthquake ground motion and some safety margin. The ground motions and surface deformation caused by capable faults are also considered in the seismic design of NPPs. In addition, the Korea Institute of Nuclear Safety as a regulatory technical expert organization, has been operating independent real time earthquake monitoring network as a part of securing the seismic safety of NPP sites in South Korea since late 1990's. If earthquakes with more than magnitude 3.0 are occurred in the Korean Peninsula or the peak ground

  18. Compositional Safety Analysis using Barrier Certificates

    DEFF Research Database (Denmark)

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

    2012-01-01

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

  19. A safety analysis of warhead balancing

    Energy Technology Data Exchange (ETDEWEB)

    Bott, T.F.

    1998-12-01

    Reentry vehicles (RVs) carrying warheads from ballistic missiles must be carefully balanced with the warhead in situ to prevent wobble as the RVs enter the earth`s atmosphere to prevent inaccuracy or loss of the warhead. This balancing is performed on a dynamic balancing machine that rotates the RV at significant angular velocities. Seizure of the spindle shaft of the machine could result in rapid deceleration of the rotating assembly, which could over-stress and shear bolts or other structures that attach the RV to the balancing machine. This could result in undesired motions of the RV and impact of the RV on equipment or structures in the work area. This potential safety problem has long been recognized in a general way, but no systematic investigation of the possible accident sequences had been performed. The purpose of this paper is to describe an integrated set of systems analysis techniques that worked well in developing a set of accident sequences that describe the motions of the RV following a spindle-shaft seizure event.

  20. Safety assessment of animal- and plant-derived amino acids as used in cosmetics.

    Science.gov (United States)

    Burnett, Christina; Heldreth, Bart; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan

    2014-01-01

    The Cosmetic Ingredient Review Expert Panel (Panel) reviewed the safety of animal- and plant-derived amino acid mixtures, which function as skin and hair conditioning agents. The safety of α-amino acids as direct food additives has been well established, based on extensive research through acute and chronic dietary exposures and the Panel previously has reviewed the safety of individual α-amino acids in cosmetics. The Panel focused its review on dermal irritation and sensitization data relevant to the use of these ingredients in topical cosmetics. The Panel concluded that these 21 ingredients are safe in the present practices of use and concentration as used in cosmetics.

  1. The Experience-Based Safety Training System Using Vr Technology for Chemical Plant

    Directory of Open Access Journals (Sweden)

    Atsuko Nakai

    2014-11-01

    Full Text Available In chemical plants, safety measures are needed in order to minimize the impact of severe accidents and natural disasters. At the same time, carrying out the education and training to workers the corresponding operation in non-stationary situation is essential. However, reproducing the non-stationary conditions to actual equipment or mock-up cannot be performed because it is dangerous. By using the virtual reality (VR technology, we can build up a virtual chemical plant with lower cost compared to real plant. The operator can experience the fire and explosion accidents in the virtual space. Therefore, in this paper, we propose an experienced-based safety training system for implementing the education and training by using the non-stationary situation in the computer. This proposed system is linked with the dynamic plant simulator. A trainee can learn the correct operation through the simulated experience to prevent an accident. The safety awareness of workers will improve by experiential learning. The proposed system is useful for safety education in chemical plant.

  2. Meta-analysis of surgical safety checklist effects on teamwork, communication, morbidity, mortality, and safety.

    Science.gov (United States)

    Lyons, Vanessa E; Popejoy, Lori L

    2014-02-01

    The purpose of this study is to examine the effectiveness of surgical safety checklists on teamwork, communication, morbidity, mortality, and compliance with safety measures through meta-analysis. Four meta-analyses were conducted on 19 studies that met the inclusion criteria. The effect size of checklists on teamwork and communication was 1.180 (p = .003), on morbidity and mortality was 0.123 (p = .003) and 0.088 (p = .001), respectively, and on compliance with safety measures was 0.268 (p teamwork and communication, reduce morbidity and mortality, and improve compliance with safety measures. This meta-analysis is limited in its generalizability based on the limited number of studies and the inclusion of only published research. Future research is needed to examine possible moderating variables for the effects of surgical safety checklists.

  3. Cost Benefit Analysis of Consumer Product Safety Standards

    Science.gov (United States)

    Smith, Betty F.; Dardis, Rachel

    1977-01-01

    This paper investigates the role of cost-benefit analysis in evaluating consumer product safety standards and applys such analysis to an evaluation of flammability standards for children's sleepwear. (Editor)

  4. Hazards and hazard combinations relevant for the safety of nuclear power plants

    Science.gov (United States)

    Decker, Kurt; Brinkman, Hans; Raimond, Emmanuel

    2017-04-01

    exclusive (e.g., extremely high air temperature and surface ice). Our dataset further provides information on hazard combinations which are more likely to occur than just by random coincidence. 577 correlations between individual hazards are identified by expert opinion and shown in a cross-correlation chart. Combinations discriminate between: (1) causally connected hazards (cause-effect relation) where one hazard (e.g., costal erosion) may be caused by another hazard (e.g., storm surge); or where one hazard (e.g., high wind) is a prerequisite for a correlated hazard (e.g., storm surge). The identified causal links are not commutative. (2) Associated hazards ("contemporary" events) which are probable to occur at the same time due to a common root cause (e.g., a cold front of a meteorological low pressure area which leads to a drop of air pressure, high wind, thunderstorm, lightning, heavy rain and hail). The root cause may not necessarily be regarded as a hazard by itself. The hazard list and the hazard correlation chart may serve as a starting point for the hazard analysis process for nuclear installations in Level 1 PSA as outlined by IAEA (2010), the definition of design basis for nuclear reactors, and the assessment of design extension conditions as required by WENRA-RHWG (2014). It may further be helpful for the identification of hazard combinations and hazard cascades which threaten other critical infrastructure. References: Decker, K. & Brinkman, H., 2017. List of external hazards to be considered in extended PSA. Report No. ASAMPSA_E/WP21/D21.2/2017-41 - IRSN/ PSN-RES/SAG/2017-00011 IAEA, 2010. Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants. Safety Guide No. SSG-3, Vienna. http://www-pub.iaea.org/books/ WENRA-RHWG, 2014. WENRA Safety Reference Levels for Existing Reactors. Update in Relation to Lessons Learned from TEPCO Fukushima Dai-Ichi Accident. http://www.wenra.org/publications/

  5. Safety analysis report 231-Z Building

    Energy Technology Data Exchange (ETDEWEB)

    Powers, C.S.

    1989-03-01

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

  6. Plant safety in France 2000 until 2010; Anlagensicherheitspolitik in Frankreich 2000 bis 2010

    Energy Technology Data Exchange (ETDEWEB)

    Vallee, Agnes; Affeltranger, Bastien; Descourriere, Sandrine; Oger, Florence; Duval, Christophe; Gaucher, Rodolphe [Institut National de l' Environnement Industriel et des Risques (INERIS), 60 - Verneuil-en-Halatte (France)

    2013-03-15

    The study on the safety of industrial plants on France covers the following issues: The explosion in the operational area of AZF (plant Grande Paroisse) and the consequences; classified installations; the players of accident risks in France; realization of the accident prevention policy in France from 2000 to 2010; prevention of natural risks and consideration of these risks on the level of classified installations (NATECH accidents) in France from 2000 to 2010; further specific topics; actual activities and developments.

  7. 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.

  8. Interrelationships of food safety and plant pathology: the life cycle of human pathogens on plants.

    Science.gov (United States)

    Barak, Jeri D; Schroeder, Brenda K

    2012-01-01

    Bacterial food-borne pathogens use plants as vectors between animal hosts, all the while following the life cycle script of plant-associated bacteria. Similar to phytobacteria, Salmonella, pathogenic Escherichia coli, and cross-domain pathogens have a foothold in agricultural production areas. The commonality of environmental contamination translates to contact with plants. Because of the chronic absence of kill steps against human pathogens for fresh produce, arrival on plants leads to persistence and the risk of human illness. Significant research progress is revealing mechanisms used by human pathogens to colonize plants and important biological interactions between and among bacteria in planta. These findings articulate the difficulty of eliminating or reducing the pathogen from plants. The plant itself may be an untapped key to clean produce. This review highlights the life of human pathogens outside an animal host, focusing on the role of plants, and illustrates areas that are ripe for future investigation.

  9. Advances in coupled safety modeling using systems analysis and high-fidelity methods.

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H.; Thomas, J. W.; Nuclear Engineering Division

    2010-05-31

    The potential for a sodium-cooled fast reactor to survive severe accident initiators with no damage has been demonstrated through whole-plant testing in EBR-II and FFTF. Analysis of the observed natural protective mechanisms suggests that they would be characteristic of a broad range of sodium-cooled fast reactors utilizing metal fuel. However, in order to demonstrate the degree to which new, advanced sodium-cooled fast reactor designs will possess these desired safety features, accurate, high-fidelity, whole-plant dynamics safety simulations will be required. One of the objectives of the advanced safety-modeling component of the Reactor IPSC is to develop a science-based advanced safety simulation capability by utilizing existing safety simulation tools coupled with emerging high-fidelity modeling capabilities in a multi-resolution approach. As part of this integration, an existing whole-plant systems analysis code has been coupled with a high-fidelity computational fluid dynamics code to assess the impact of high-fidelity simulations on safety-related performance. With the coupled capabilities, it is possible to identify critical safety-related phenomenon in advanced reactor designs that cannot be resolved with existing tools. In this report, the impact of coupling is demonstrated by evaluating the conditions of outlet plenum thermal stratification during a protected loss of flow transient. Outlet plenum stratification was anticipated to alter core temperatures and flows predicted during natural circulation conditions. This effect was observed during the simulations. What was not anticipated, however, is the far-reaching impact that resolving thermal stratification has on the whole plant. The high temperatures predicted at the IHX inlet due to thermal stratification in the outlet plenum forces heat into the intermediate system to the point that it eventually becomes a source of heat for the primary system. The results also suggest that flow stagnation in the

  10. Status of safety issues at licensed power plants: TMI Action Plan requirements; unresolved safety issues; generic safety issues; other multiplant action issues. Supplement 3

    Energy Technology Data Exchange (ETDEWEB)

    1993-12-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. This third annual NUREG report, Supplement 3, presents updated information as of September 30, 1993. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ``A Prioritization of Generic Safety Issues,`` which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees.

  11. The quality/safety medical index: implementation and analysis.

    Science.gov (United States)

    Reiner, Bruce I

    2015-02-01

    Medical analytics relating to quality and safety measures have become particularly timely and of high importance in contemporary medical practice. In medical imaging, the dynamic relationship between medical imaging quality and radiation safety creates challenges in quantifying quality or safety independently. By creating a standardized measurement which simultaneously accounts for quality and safety measures (i.e., quality safety index), one can in theory create a standardized method for combined quality and safety analysis, which in turn can be analyzed in the context of individual patient, exam, and clinical profiles. The derived index measures can be entered into a centralized database, which in turn can be used for comparative performance of individual and institutional service providers. In addition, data analytics can be used to create customizable educational resources for providers and patients, clinical decision support tools, technology performance analysis, and clinical/economic outcomes research.

  12. Pinellas Plant: Child Care/Partnership School safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-11-01

    The Albuquerque Operations Office through the Pinellas Plant Area Office is involved in a joint venture to establish a Partnership School and a Day Care Facility at the Plant. The venture is unique in that it is based on a partnership with the local county school system. The county school system will provide the teachers, supplies and classroom furnishings for the operation of the school for pre-kindergarten, kindergarten, first and second grade during regular school hours. The Government will provide the facility and its normal operating and maintenance costs. A Day Care Facility will also be available for children from infancy through the second grade for outside school hours. The day care will be operated as a non-profit corporation. Fees paid by parents with children in the day care center will cove the cost of staff, food, supplies and liability insurance. Again, the government will provide the facility and its normal operating and maintenance costs. Between 75 and 90 children are expected in the first year of operation. The Partnership School will consist of one class each for pre-kindergarten, kindergarten and first grade. Second grade will be added in 1990. The total estimated number of children for both the Child Care and Partnership School should not exceed 200 children. Expected benefits include reduced absenteeism, tardiness and turnover and thus increased productivity. The program will be an asset in recruiting and retaining the best workforce. Other benefits include improved education for the children.

  13. Safety and reliability analysis in a polyvinyl chloride batch process using dynamic simulator-case study: Loss of containment incident.

    Science.gov (United States)

    Rizal, Datu; Tani, Shinichi; Nishiyama, Kimitoshi; Suzuki, Kazuhiko

    2006-10-11

    In this paper, a novel methodology in batch plant safety and reliability analysis is proposed using a dynamic simulator. A batch process involving several safety objects (e.g. sensors, controller, valves, etc.) is activated during the operational stage. The performance of the safety objects is evaluated by the dynamic simulation and a fault propagation model is generated. By using the fault propagation model, an improved fault tree analysis (FTA) method using switching signal mode (SSM) is developed for estimating the probability of failures. The timely dependent failures can be considered as unavailability of safety objects that can cause the accidents in a plant. Finally, the rank of safety object is formulated as performance index (PI) and can be estimated using the importance measures. PI shows the prioritization of safety objects that should be investigated for safety improvement program in the plants. The output of this method can be used for optimal policy in safety object improvement and maintenance. The dynamic simulator was constructed using Visual Modeler (VM, the plant simulator, developed by Omega Simulation Corp., Japan). A case study is focused on the loss of containment (LOC) incident at polyvinyl chloride (PVC) batch process which is consumed the hazardous material, vinyl chloride monomer (VCM).

  14. Applying importance-performance analysis to patient safety culture.

    Science.gov (United States)

    Lee, Yii-Ching; Wu, Hsin-Hung; Hsieh, Wan-Lin; Weng, Shao-Jen; Hsieh, Liang-Po; Huang, Chih-Hsuan

    2015-01-01

    The Sexton et al.'s (2006) safety attitudes questionnaire (SAQ) has been widely used to assess staff's attitudes towards patient safety in healthcare organizations. However, to date there have been few studies that discuss the perceptions of patient safety both from hospital staff and upper management. The purpose of this paper is to improve and to develop better strategies regarding patient safety in healthcare organizations. The Chinese version of SAQ based on the Taiwan Joint Commission on Hospital Accreditation is used to evaluate the perceptions of hospital staff. The current study then lies in applying importance-performance analysis technique to identify the major strengths and weaknesses of the safety culture. The results show that teamwork climate, safety climate, job satisfaction, stress recognition and working conditions are major strengths and should be maintained in order to provide a better patient safety culture. On the contrary, perceptions of management and hospital handoffs and transitions are important weaknesses and should be improved immediately. Research limitations/implications - The research is restricted in generalizability. The assessment of hospital staff in patient safety culture is physicians and registered nurses. It would be interesting to further evaluate other staff's (e.g. technicians, pharmacists and others) opinions regarding patient safety culture in the hospital. Few studies have clearly evaluated the perceptions of healthcare organization management regarding patient safety culture. Healthcare managers enable to take more effective actions to improve the level of patient safety by investigating key characteristics (either strengths or weaknesses) that healthcare organizations should focus on.

  15. Use of probabilistic safety analysis for design of emergency mitigation systems in hydrogen producer plant with sulfur-iodine technology, Section II: sulfuric acid decomposition; Uso de analisis probabilistico de seguridad para el diseno de sistemas de mitigacion de emergencia en planta productora de hidrogeno con tecnologia azufre-iodo, Seccion II: descomposicion de acido sulfurico

    Energy Technology Data Exchange (ETDEWEB)

    Mendoza A, A.; Nelson E, P. F.; Francois L, J. L. [Facultad de Ingenieria, Departamento de Sistemas Energeticos, UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)], e-mail: iqalexmdz@yahoo.com.mx

    2009-10-15

    Over the last decades, the need to reduce emissions of greenhouse gases has prompted the development of technologies for the production of clean fuels through the use of primary energy resources of zero emissions, as the heat of nuclear reactors of high temperature. Within these technologies, one of the most promising is the hydrogen production by sulfur-iodine cycle coupled to a high temperature reactor initially proposed by General Atomics. By their nature and because it will be large-scale plants, the development of these technologies from its present phase to its procurement and construction, will have to incorporate emergency mitigation systems in all its parts and interconnections to prevent undesired events that could put threaten the plant integrity and the nearby area. For the particular case of sulfur-iodine thermochemical cycle, most analysis have focused on hydrogen explosions and failures in the primary cooling systems. While these events are the most catastrophic, is that there are also many other events that even taking less direct consequences, could jeopardize the plant operation, the people safety of nearby communities and carry the same economic consequences. In this study we analyzed one of these events, which is the formation of a toxic cloud prompted by uncontrolled leakage of concentrated sulfuric acid in the second section of sulfur-iodine process of General Atomics. In this section, the sulfuric acid concentration is near to 90% in conditions of high temperature and positive pressure. Under these conditions the sulfuric acid and sulfur oxides from the reactor will form a toxic cloud that the have contact with the plant personnel could cause fatalities, or to reach a town would cause suffocation, respiratory problems and eye irritation. The methodology used for this study is the supported design in probabilistic safety analysis. Mitigation systems were postulated based on the isolation of a possible leak, the neutralization of a pond of

  16. Mathematical aspects of assessing extreme events for the safety of nuclear plants

    Science.gov (United States)

    Potempski, Slawomir; Borysiewicz, Mieczyslaw

    2015-04-01

    In the paper the review of mathematical methodologies applied for assessing low frequencies of rare natural events like earthquakes, tsunamis, hurricanes or tornadoes, floods (in particular flash floods and surge storms), lightning, solar flares, etc., will be given in the perspective of the safety assessment of nuclear plants. The statistical methods are usually based on the extreme value theory, which deals with the analysis of extreme deviation from the median (or the mean). In this respect application of various mathematical tools can be useful, like: the extreme value theorem of Fisher-Tippett-Gnedenko leading to possible choices of general extreme value distributions, or the Pickands-Balkema-de Haan theorem for tail fitting, or the methods related to large deviation theory. In the paper the most important stochastic distributions relevant for performing rare events statistical analysis will be presented. This concerns, for example, the analysis of the data with the annual extreme values (maxima - "Annual Maxima Series" or minima), or the peak values, exceeding given thresholds at some periods of interest ("Peak Over Threshold"), or the estimation of the size of exceedance. Despite of the fact that there is a lack of sufficient statistical data directly containing rare events, in some cases it is still possible to extract useful information from existing larger data sets. As an example one can consider some data sets available from the web sites for floods, earthquakes or generally natural hazards. Some aspects of such data sets will be also presented taking into account their usefulness for the practical assessment of risk for nuclear power plants coming from extreme weather conditions.

  17. 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.

  18. 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.

  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. Improved technical specifications and related improvements to safety in commercial Nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, D.R.; Demitrack, T.; Schiele, R.; Jones, J.C. [EXCEL Services Corporation, 11921 Rockville Pike, Suite 100, Rockville, MD 20852 (United States)]. e-mail: donaldh@excelservices.com

    2004-07-01

    Many of the commercial nuclear power plants in the United States (US) have been converting a portion of the plant operating license known as the Technical Specifications (TS) in accordance with a document published by the US Nuclear Regulatory Commission (NRC). The TS prescribe commercial nuclear power plant operating requirements. There are several types of nuclear power plants in the US, based on the technology of different vendors, and there is an NRC document that supports each of the five different vendor designs. The NRC documents are known as the Improved Standard Technical Specifications (ISTS) and are contained in a separate document (NUREG series) for each one of the designs. EXCEL Services Corporation (hereinafter EXCEL) has played a major role in the development of the ISTS and in the development, licensing, and implementation of the plant specific Improved Technical Specifications (ITS) (which is based on the ISTS) for the commercial nuclear power plants in the US that have elected to make this conversion. There are currently 103 operating commercial nuclear power plants in the US and 68 of them have successfully completed the conversion to the ITS and are now operating in accordance with their plant specific ITS. The ISTS is focused mainly on safety by ensuring the commercial nuclear reactors can safely shut down and mitigate the consequences of any postulated transient and accident. It accomplishes this function by including requirements directly associated with safety in a document structured systematically and taking into account some key human factors and technical initiatives. This paper discusses the ISTS including its format, content, and detail, the history of the ISTS, the ITS development, licensing, and implementation process, the safety improvements resulting from a plant conversion to ITS, and the importance of the ITS Project to the industry. (Author)

  1. Technical safety appraisal: Buildings 776/777 Rocky Flats Plant

    Energy Technology Data Exchange (ETDEWEB)

    Field, H C

    1988-03-01

    Buildings 776/777 at the Rocky Flats Plant are major components of the production complex at the plant site. They have been in operation since 1957. The operations taking place in the buildings are nuclear weapons production support, processing of weapons assemblies returned from Pantex, waste processing, research and development in support of production, special projects, and those generated by support groups, such as maintenance. The appraisal team identified nine deficiencies that it believed required prompt attention. DOE management for EH, the program office (Defense Programs), and the field office analyzed the information provided by the appraisal team and instituted compensatory measures for closer monitoring of contractor activities by knowledgeable DOE staff and staff from other sites. Concurrently, the contractor was requested to address both short-term and long-term remedial measures to correct the identified issues as well as the underlying problems. The contractor has provided his action plan, which is included. This plan was under evaluation by EH and the DOE program office at the time this report was prepared. In addressing the major areas of concern identified above, a well as the specific deficiencies identified by the appraisal team, the contractor and the field office are cautioned to search for the root causes for the problems and to direct corrective actions to those root causes rather than solely to the symptoms to assure the sustainability of the improvements being made. The results of prior TSAs led DOE to conclude that previous corrective actions were not sufficient in that a large number of the individual findings are recurrent. Pending completion of remedial actions over the next few months, enhanced DOE oversight of the contractor is warranted.

  2. Next Generation Nuclear Plant GAP Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Sydney J [ORNL; Burchell, Timothy D [ORNL; Corwin, William R [ORNL; Fisher, Stephen Eugene [ORNL; Forsberg, Charles W. [Massachusetts Institute of Technology (MIT); Morris, Robert Noel [ORNL; Moses, David Lewis [ORNL

    2008-12-01

    As a follow-up to the phenomena identification and ranking table (PIRT) studies conducted recently by NRC on next generation nuclear plant (NGNP) safety, a study was conducted to identify the significant 'gaps' between what is needed and what is already available to adequately assess NGNP safety characteristics. The PIRT studies focused on identifying important phenomena affecting NGNP plant behavior, while the gap study gives more attention to off-normal behavior, uncertainties, and event probabilities under both normal operation and postulated accident conditions. Hence, this process also involved incorporating more detailed evaluations of accident sequences and risk assessments. This study considers thermal-fluid and neutronic behavior under both normal and postulated accident conditions, fission product transport (FPT), high-temperature metals, and graphite behavior and their effects on safety. In addition, safety issues related to coupling process heat (hydrogen production) systems to the reactor are addressed, given the limited design information currently available. Recommendations for further study, including analytical methods development and experimental needs, are presented as appropriate in each of these areas.

  3. Next Generation Nuclear Plant Structures, Systems, and Components Safety Classification White Paper

    Energy Technology Data Exchange (ETDEWEB)

    Pete Jordan

    2010-09-01

    This white paper outlines the relevant regulatory policy and guidance for a risk-informed approach for establishing the safety classification of Structures, Systems, and Components (SSCs) for the Next Generation Nuclear Plant and sets forth certain facts for review and discussion in order facilitate an effective submittal leading to an NGNP Combined Operating License application under 10 CFR 52.

  4. Artificial neural network model for prediction of safety performance indicators goals in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Souto, Kelling C.; Nunes, Wallace W. [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro, Nilopolis, RJ (Brazil). Lab. de Aplicacoes Computacionais; Machado, Marcelo D., E-mail: dornemd@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil). Gerencia de Combustivel Nuclear - GCN.T

    2011-07-01

    Safety performance indicators have been developed to provide a quantitative indication of the performance and safety in various industry sectors. These indexes can provide assess to aspects ranging from production, design, and human performance up to management issues in accordance with policy, objectives and goals of the company. The use of safety performance indicators in nuclear power plants around the world is a reality. However, it is necessary to periodically set goal values. Such goals are targets relating to each of the indicators to be achieved by the plant over a predetermined period of operation. The current process of defining these goals is carried out by experts in a subjective way, based on actual data from the plant, and comparison with global indices. Artificial neural networks are computational techniques that present a mathematical model inspired by the neural structure of intelligent organisms that acquire knowledge through experience. This paper proposes an artificial neural network model aimed at predicting values of goals to be used in the evaluation of safety performance indicators for nuclear power plants. (author)

  5. Safety of long-term consumption of plant sterol esters-enriched spread

    NARCIS (Netherlands)

    Hendriks, H.F.J.; Brink, E.J.; Meijer, G.W.; Princen, H.M.G.; Ntanios, F.Y.

    2003-01-01

    Objective: To evaluate both efficacy and safety in humans of long-term consumption of spreads containing plant sterol esters. Design: Randomized double-blind placebo-controlled parallel trial. Subjects: Hundred and eighty-five healthy volunteers (35-64y). Intervention: Volunteers daily consumed 20g

  6. Artificial neural network model for prediction of safety performance indicators goals in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Souto, Kelling C.; Nunes, Wallace W. [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro, Nilopolis, RJ (Brazil). Lab. de Aplicacoes Computacionais; Machado, Marcelo D., E-mail: dornemd@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil). Gerencia de Combustivel Nuclear - GCN.T

    2011-07-01

    Safety performance indicators have been developed to provide a quantitative indication of the performance and safety in various industry sectors. These indexes can provide assess to aspects ranging from production, design, and human performance up to management issues in accordance with policy, objectives and goals of the company. The use of safety performance indicators in nuclear power plants around the world is a reality. However, it is necessary to periodically set goal values. Such goals are targets relating to each of the indicators to be achieved by the plant over a predetermined period of operation. The current process of defining these goals is carried out by experts in a subjective way, based on actual data from the plant, and comparison with global indices. Artificial neural networks are computational techniques that present a mathematical model inspired by the neural structure of intelligent organisms that acquire knowledge through experience. This paper proposes an artificial neural network model aimed at predicting values of goals to be used in the evaluation of safety performance indicators for nuclear power plants. (author)

  7. 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.

  8. Discussion on software aging management of nuclear power plant safety digital control system.

    Science.gov (United States)

    Liang, Huihui; Gu, Pengfei; Tang, Jianzhong; Chen, Weihua; Gao, Feng

    2016-01-01

    Managing the aging of digital control systems ensures that nuclear power plant systems are in adequate safety margins during their life cycles. Software is a core component in the execution of control logic and differs between digital and analog control systems. The hardware aging management for the digital control system is similar to that for the analog system, which has matured over decades of study. However, software aging management is still in the exploratory stage. Software aging evaluation is critical given the higher reliability and safety requirements of nuclear power plants. To ensure effective inputs for reliability assessment, this paper provides the required software aging information during the life cycle. Moreover, the software aging management scheme for safety digital control system is proposed on the basis of collected aging information.

  9. Schedulability analysis of SCOPS on a platform for safety I and C systems of SMART MMIS

    Energy Technology Data Exchange (ETDEWEB)

    Keum, Jong Yong; Suh, Yong Suk; Jeong, Kwang Il; Park, Je Yun [KAERI, Daejeon (Korea, Republic of); Seo, Yong Jin; Kim, Hyeon Soo [Nat' l Univ., Daejon (Korea, Republic of)

    2012-10-15

    A real time I and C system used in safety systems in nuclear power plants shall have predictable and deterministic characteristics. The main issue of predictable real time system is to prove whether it satisfies its deadline. One way to prove whether a realtime I and C system satisfies its deadline is a schedulability analysis. A schedulability analysis on SCOPS (SMART Core Protection System) is performed.

  10. Analysis on Pollution Factors in Asparagus Production and Research on Safety Production Technology

    OpenAIRE

    Ma, Liping; Hao, Bianqing; Qiao, Xiongwu

    2013-01-01

    Based on the analysis on the infection degree, infection law and influencing factors of the main diseases on asparagus and the analysis on the pollution factors in asparagus production such as blind pesticide use, atmospheric pollution and acid rain, the pollution of soil and fertilizer, this article proposes asparagus safety production technologies which include the selection of disease-resistant variety and suitable planting field, scientific and reasonable disease control, balanced fertili...

  11. Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Su, T.M.

    1981-05-01

    Inplant tests of safety/relief valve (SRV) discharges may be required to confirm generically established specifications for SRV loads and the maximum suppression pool temperature, and to evaluate possible effects of plant-unique parameters. These tests are required in those plants which have features that differ substantially from those previously tested. Guidelines for formulating appropriate test matrices, establishing test procedures, selecting necessary instrumentation, and reporting the test results are provided in this report. Guidelines to determine if inplant tests are required on the basis of the plant unique parameters are also included in the report.

  12. Vulnerability, safety and response of nuclear power plants to the hydroclimatic hazards

    Science.gov (United States)

    János Katona, Tamás; Vilimi, András

    2016-04-01

    The Great Tohoku Earthquake and Tsunami, and the severe accident at Fukushima Dai-ichi nuclear power plant 2011 alerted the nuclear industry to danger of extreme rare natural hazards. The subsequent "stress tests" performed by the nuclear industry in Europe and all over the world identifies the nuclear power plant (NPP) vulnerabilities and define the measures for increasing the plant safety. According to the international practice of nuclear safety regulations, the cumulative core damage frequency for NPPs has to be 10-5/a, and the cumulative frequency of early large release has to be 10-6/a. In case of operating plants these annual probabilities can be little higher, but the licensees are obliged to implement all reasonable practicable measures for increasing the plant safety. For achieving the required level of safety, design basis of NPPs for natural hazards has to be defined at the 10-4/a ⎯10-5/a levels of annual exceedance probability. Tornado hazard is some kind of exception, e.g., the design basis annual probability for tornado in the US is equal to 10-7/a. Design of the NPPs shall provide for an adequate margin to protect items ultimately necessary to prevent large or early radioactive releases in the event of levels of natural hazards exceeding those to be considered for design. The plant safety has to be reviewed for accounting the changes of the environmental conditions and natural hazards in case of necessity, but as minimum every ten years in the frame of periodic safety reviews. Long-term forecast of environmental conditions and hazards has to be accounted for in the design basis of the new plants. Changes in hydroclimatic variables, e.g., storms, tornadoes, river floods, flash floods, extreme temperatures, droughts affect the operability and efficiency as well as the safety the NPPs. Low flow rates and high water temperature in the rivers may force to operate at reduced power level or shutdown the plant (Cernavoda NPP, Romania, August 2009). The

  13. Studies of the relationship between employee`s safety consciousness, morale, and supervisor`s leadership in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Misumi, Jyuji; Hiraki, Tadao; Sakurai, Yukihiro [Institute of Nuclear Safety System Inc., Kyoto (Japan); Yoshida, Michio; Misumi, Emiko; Tokudome, Eiji

    1996-09-01

    This study examined the relationship between employee`s safety consciousness, morale, and supervisor`s leadership using multiple regression analysis. Respondents were 2152 male employees who were working at nuclear power plants (operation division, maintenance division, and joint companies). Main results were as follows. (1) Individual morale variables, such as `work motivation` and `mental hygine`, were correlated with leadership M behavior rather than with P behavior. On the other hand, group morale variables, such as `teamwork` and `meeting quality`, were correlated with both P and M behavior. These results shows P and M leadership affect the employee`s morale. (2) With regard to safety consciousness variables, `communication` and `work place norm` to ensure safety were strongly correlated to leadership both P and M behavior. However, neither `sense of tension to ensure safety` nor `experiencing cold shiver` were related to leadership P or M behavior. It was suggested that practices for accidents prevention in workplace are related to supervisor`s P and M leadership behavior. (3) `Sense of tension` to ensure safety and `experiencing cold shiver` were negatively correlated with `mental hygine`, but positively correlated with `work motivation`. These results suggest that increase of the work motivation might improve employee`s awareness and ability for detecting human errors. (author)

  14. PROBABILISTIC SAFETY ASSESSMENT OF OPERATIONAL ACCIDENTS AT THE WASTE ISOLATION PILOT PLANT

    Energy Technology Data Exchange (ETDEWEB)

    Rucker, D.F.

    2000-09-01

    This report presents a probabilistic safety assessment of radioactive doses as consequences from accident scenarios to complement the deterministic assessment presented in the Waste Isolation Pilot Plant (WIPP) Safety Analysis Report (SAR). The International Council of Radiation Protection (ICRP) recommends both assessments be conducted to ensure that ''an adequate level of safety has been achieved and that no major contributors to risk are overlooked'' (ICRP 1993). To that end, the probabilistic assessment for the WIPP accident scenarios addresses the wide range of assumptions, e.g. the range of values representing the radioactive source of an accident, that could possibly have been overlooked by the SAR. Routine releases of radionuclides from the WIPP repository to the environment during the waste emplacement operations are expected to be essentially zero. In contrast, potential accidental releases from postulated accident scenarios during waste handling and emplacement could be substantial, which necessitates the need for radiological air monitoring and confinement barriers (DOE 1999). The WIPP Safety Analysis Report (SAR) calculated doses from accidental releases to the on-site (at 100 m from the source) and off-site (at the Exclusive Use Boundary and Site Boundary) public by a deterministic approach. This approach, as demonstrated in the SAR, uses single-point values of key parameters to assess the 50-year, whole-body committed effective dose equivalent (CEDE). The basic assumptions used in the SAR to formulate the CEDE are retained for this report's probabilistic assessment. However, for the probabilistic assessment, single-point parameter values were replaced with probability density functions (PDF) and were sampled over an expected range. Monte Carlo simulations were run, in which 10,000 iterations were performed by randomly selecting one value for each parameter and calculating the dose. Statistical information was then derived

  15. 78 FR 47011 - Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2013-08-02

    ... COMMISSION Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants..., ``Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.'' This... software elements if those systems include software. This RG is one of six RG revisions addressing...

  16. Safety analysis of passing maneuvers using extreme value theory

    Directory of Open Access Journals (Sweden)

    Haneen Farah

    2017-04-01

    The results indicate that this is a promising approach for safety evaluation. On-going work of the authors will attempt to generalize this method to other safety measures related to passing maneuvers, test it for the detailed analysis of the effect of demographic factors on passing maneuvers' crash probability and for its usefulness in a traffic simulation environment.

  17. 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.

  18. 77 FR 50722 - Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2012-08-22

    ... COMMISSION Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants... regulatory guide (DG), DG-1208, ``Software Unit Testing for Digital Computer Software used in Safety Systems... entitled ``Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear...

  19. Systems Analysis of NASA Aviation Safety Program: Final Report

    Science.gov (United States)

    Jones, Sharon M.; Reveley, Mary S.; Withrow, Colleen A.; Evans, Joni K.; Barr, Lawrence; Leone, Karen

    2013-01-01

    A three-month study (February to April 2010) of the NASA Aviation Safety (AvSafe) program was conducted. This study comprised three components: (1) a statistical analysis of currently available civilian subsonic aircraft data from the National Transportation Safety Board (NTSB), the Federal Aviation Administration (FAA), and the Aviation Safety Information Analysis and Sharing (ASIAS) system to identify any significant or overlooked aviation safety issues; (2) a high-level qualitative identification of future safety risks, with an assessment of the potential impact of the NASA AvSafe research on the National Airspace System (NAS) based on these risks; and (3) a detailed, top-down analysis of the NASA AvSafe program using an established and peer-reviewed systems analysis methodology. The statistical analysis identified the top aviation "tall poles" based on NTSB accident and FAA incident data from 1997 to 2006. A separate examination of medical helicopter accidents in the United States was also conducted. Multiple external sources were used to develop a compilation of ten "tall poles" in future safety issues/risks. The top-down analysis of the AvSafe was conducted by using a modification of the Gibson methodology. Of the 17 challenging safety issues that were identified, 11 were directly addressed by the AvSafe program research portfolio.

  20. Application of Safety Instrumented System (SIS) approach in older nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca

    2016-05-15

    Highlights: • Study Safety Instrumented System (SIS) design for older nuclear power plant. • Apply SIS on Reheater Drains (RD) system. • Apply IEC 61508/61511 to design safety system. • Evaluate risk reduction based on proposed SIS design. - Abstract: In order to remain economically effective and financially profitable, the modern industries have to take their safety culture to a higher level and consider production losses in addition to simple accident prevention techniques. Ideally, compliance with safety requirements start during early design stages, but in some older facilities provisions for Safety Instrumented Systems (SIS) may not have been originally included. In this paper, a case study of a Reheater Drains (RD) system is used to illustrate such an example. Frequent failures of tank level controller lead to transients where the operation of shutting down RD pumps requires operators to manually isolate the quenching water and to close the main steam admission valves. Water in this system is at saturation temperature for the reheater steam side pressure, and any manual operation of the system is highly undesirable due to hazards of working with wet steam at approximately 758 kPa(g) pressure, preheated to 237 °C. Additionally, losses of inventory are highly undesirable as well and challenge other systems in the plant. In this paper, it is suggested that RD system can benefit from installation of an independent SIS system in order to address current challenges. This idea is being explored using IEC 61508 framework for “Functional safety of electrical/electronic/programmable electronic safety-related systems” to provide assurance that the SIS will offer the necessary risk reduction required to achieve required safety for the equipment.

  1. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2001-06-01

    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. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER 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 benign 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. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are 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 energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  2. Safety analysis for key design features of KALIMER-600 design concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S

    2005-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign 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 using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed.

  3. An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, Roger J.

    1989-09-01

    This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.

  4. An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, Roger J.

    1989-09-01

    This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.

  5. 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...

  6. Computational Analysis of Safety Injection Tank Performance

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Oan; Nietiadia, Yohanes Setiawan; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Addad, Yacine; Yoon, Ho Joon [Khalifa University of Science Technology and Research, Abu Dhabi (United Arab Emirates)

    2015-10-15

    The APR 1400 is a large pressurized water reactor (PWR). Just like many other water reactors, it has an emergency core cooling system (ECCS). One of the most important components in the ECCS is the safety injection tank (SIT). Inside the SIT, a fluidic device is installed, which passively controls the mass flow of the safety injection and eliminates the need for low pressure safety injection pumps. As more passive safety mechanisms are being pursued, it has become more important to understand flow structure and the loss mechanism within the fluidic device. Current computational fluid dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study proposes to find a more exact result using CFD and more realistic modeling. The SIT of APR1400 was analyzed using MARS and CFD. CFD calculation was executed first to obtain the form loss factor. Using the two form loss factors from the vendor and calculation, calculation using MARS was performed to compare with experiment. The accumulator model in MARS was quite accurate in predicting the water level. The pipe model showed some difference with the experimental data in the water level.

  7. Preliminary safety design analysis of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

  8. An analysis of the traffic safety phenomenon.

    NARCIS (Netherlands)

    Asmussen, E. & Kranenburg, A.

    1982-01-01

    The lack of traffic safety is a combination of the critical coincidence of circumstances in the traffic of incidents (near-accidents) and accidents with unwanted (permanent) consequences, such as fatalities, injured and disabled persons and material damage. This definition covers the whole of the cr

  9. Analysis of Main Steam Safety Valve Improvement for Qinshan Nuclear Power Plant Phase 2 Unit 1 & 2%秦山第二核电厂1 & 2号机组主蒸汽安全阀改进分析

    Institute of Scientific and Technical Information of China (English)

    任春明; 陈坚刚; 黄代顺

    2015-01-01

    秦山第二核电厂1&2号机组拟将其加能助动式主蒸汽安全阀改为弹簧加载式安全阀。通过比对秦山第二核电厂3&4号机组主蒸汽安全阀设计,提出了1&2号机组主蒸汽安全阀改进方案,即第1组阀门采用弹簧加载式并调整开启整定值,并从机械设计、仪控设计和安全分析等方面论证了该方案的可行性。新的改进方案在保证安全的前提下,简化了设计,大幅减少了工程投入,同时降低了系统和控制逻辑复杂化后带来的潜在停堆风险的增加。%An improvement for Qinshan Nuclear Power Plant Phase 2 unit 1&2 ,in which the power‐operated valves are replaced with the spring‐loaded valve with setpoint adjustment for the 1st group safety valves of steam generators ,would be adpoted on the basis of comparison with similar improvement implemented in Qinshan Nuclear Power Plant Phase 2 unit 3&4 .The feasibility was evaluated from several aspects ,including mechanical design ,instrument & control design and safety analysis .On the premise of guaranteeing safety ,this improvement can simplify design ,observably reduce project costs and avoid increasing the potential probability of the reactor trip risk induced by complication of the system and control logic .

  10. 76 FR 28336 - Domestic Licensing of Source Material-Amendments/Integrated Safety Analysis

    Science.gov (United States)

    2011-05-17

    ..., staff document entitled ``A Comparison of Integrated Safety Analysis and Probabilistic Risk Assessment... Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities... Safety Analysis AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The U.S....

  11. 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.

  12. 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...

  13. Semi-quantitative study to evaluate the performance of a HACCP-based food safety management system in Japanese milk processing plants

    NARCIS (Netherlands)

    Sampers, I.; Toyofuku, H.; Luning, P.A.; Uyttendaele, M.; Jacxsens, L.

    2012-01-01

    This study aimed to gain an insight in the performance of Hazard Analysis and Critical Control Points (HACCP)-based food safety management systems (FSMS) implemented in Japanese milk processing plants. Since 1995, Japan has a comprehensive approval system for food manufacturing establishments by

  14. Semi-quantitative study to evaluate the performance of a HACCP-based food safety management system in Japanese milk processing plants

    NARCIS (Netherlands)

    Sampers, I.; Toyofuku, H.; Luning, P.A.; Uyttendaele, M.; Jacxsens, L.

    2012-01-01

    This study aimed to gain an insight in the performance of Hazard Analysis and Critical Control Points (HACCP)-based food safety management systems (FSMS) implemented in Japanese milk processing plants. Since 1995, Japan has a comprehensive approval system for food manufacturing establishments by eva

  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. 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...

  17. A decade of plant proteomics and mass spectrometry: translation of technical advancements to food security and safety issues.

    Science.gov (United States)

    Agrawal, Ganesh Kumar; Sarkar, Abhijit; Righetti, Pier Giorgio; Pedreschi, Romina; Carpentier, Sebastien; Wang, Tai; Barkla, Bronwyn J; Kohli, Ajay; Ndimba, Bongani Kaiser; Bykova, Natalia V; Rampitsch, Christof; Zolla, Lello; Rafudeen, Mohamed Suhail; Cramer, Rainer; Bindschedler, Laurence Veronique; Tsakirpaloglou, Nikolaos; Ndimba, Roya Janeen; Farrant, Jill M; Renaut, Jenny; Job, Dominique; Kikuchi, Shoshi; Rakwal, Randeep

    2013-01-01

    Tremendous progress in plant proteomics driven by mass spectrometry (MS) techniques has been made since 2000 when few proteomics reports were published and plant proteomics was in its infancy. These achievements include the refinement of existing techniques and the search for new techniques to address food security, safety, and health issues. It is projected that in 2050, the world's population will reach 9-12 billion people demanding a food production increase of 34-70% (FAO, 2009) from today's food production. Provision of food in a sustainable and environmentally committed manner for such a demand without threatening natural resources, requires that agricultural production increases significantly and that postharvest handling and food manufacturing systems become more efficient requiring lower energy expenditure, a decrease in postharvest losses, less waste generation and food with longer shelf life. There is also a need to look for alternative protein sources to animal based (i.e., plant based) to be able to fulfill the increase in protein demands by 2050. Thus, plant biology has a critical role to play as a science capable of addressing such challenges. In this review, we discuss proteomics especially MS, as a platform, being utilized in plant biology research for the past 10 years having the potential to expedite the process of understanding plant biology for human benefits. The increasing application of proteomics technologies in food security, analysis, and safety is emphasized in this review. But, we are aware that no unique approach/technology is capable to address the global food issues. Proteomics-generated information/resources must be integrated and correlated with other omics-based approaches, information, and conventional programs to ensure sufficient food and resources for human development now and in the future.

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

    Energy Technology Data Exchange (ETDEWEB)

    Busche, D.M.

    1995-09-01

    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.

  19. 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.

  20. The Factors Analysis on Food Safety Accidents Statistics

    Directory of Open Access Journals (Sweden)

    Boyi Xiang

    2015-05-01

    Full Text Available The study uses SPSS17.0 analysis of validity and reliability of the food enterprises questionnaire. Using AMOS17. 0 software for structural equation model test of goodness of fit and analysis of on the path. From the “melamine” to “Sudanred” and “steroid-tainted pork” events that have been exposed recently, series of typical food safety incidents resulted in the emergence of food safety issues become the focus of attention. A series of food processing can be contaminated by harmful substances, resulting in harmful food, thus constituting food safety issues and poses a serious threat to public and person’s health.

  1. Recent Progresses in Nanobiosensing for Food Safety Analysis

    Directory of Open Access Journals (Sweden)

    Tao Yang

    2016-07-01

    Full Text Available 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.

  2. The Barselina Project Phase 4 Summary report. Ignalina Unit 2 Probabilistic Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, Gunnar [ES-Konsult AB, Stockholm (Sweden); Hellstroem, P. [RELCON AB, Solna (Sweden); Zheltobriuch, G.; Bagdonas, A. [Ignalina Power Plant, Visaginas (Lithuania)

    1996-12-01

    The Barselina Project was initiated in the summer of 1991. The project is a multilateral co-operation between Lithuania, Russia and Sweden. The long range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. The Swedish BWR Barsebaeck is used as reference plant and the Lithuanian RBMK Ignalina as application plant. During phase 3, from March, 1993 to June, 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 (INPP-2) was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. To increase the realism of the risk model a set of deterministic analyses were performed and plant/RBMK-specific data bases were developed and used. A general concept for analysing this type of reactor was developed. During phase 4, July 1994 to September 1996, the PSA was further developed, taking into account plant changes, improved modeling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The updated model is quantified and new results and conclusions are evaluated.

  3. CONFERENCE REPORT: Summary of the 8th IAEA Technical Meeting on Fusion Power Plant Safety

    Science.gov (United States)

    Girard, J. Ph.; Gulden, W.; Kolbasov, B.; Louzeiro-Malaquias, A.-J.; Petti, D.; Rodriguez-Rodrigo, L.

    2008-01-01

    Reports were presented covering a selection of topics on the safety of fusion power plants. These included a review on licensing studies developed for ITER site preparation surveying common and non-common issues (i.e. site dependent) as lessons to a broader approach for fusion power plant safety. Several fusion power plant models, spanning from accessible technology to more advanced-materials based concepts, were discussed. On the topic related to fusion-specific technology, safety studies were reported on different concepts of breeding blanket modules, tritium handling and auxiliary systems under normal and accident scenarios' operation. The testing of power plant relevant technology in ITER was also assessed in terms of normal operation and accident scenarios, and occupational doses and radioactive releases under these testings have been determined. Other specific safety issues for fusion have also been discussed such as availability and reliability of fusion power plants, dust and tritium inventories and component failure databases. This study reveals that the environmental impact of fusion power plants can be minimized through a proper selection of low activation materials and using recycling technology helping to reduce waste volume and potentially open the route for its reutilization for the nuclear sector or even its clearance into the commercial circuit. Computational codes for fusion safety have been presented in support of the many studies reported. The on-going work on establishing validation approaches aiming at improving the prediction capability of fusion codes has been supported by experimental results and new directions for development have been identified. Fusion standards are not available and fission experience is mostly used as the framework basis for licensing and target design for safe operation and occupational and environmental constraints. It has been argued that fusion can benefit if a specific fusion approach is implemented, in particular

  4. Periodical safety review of the Goesgen-Daeniken nuclear power plant. Summary, results and evaluation; Periodische Sicherheitsueberpruefung fuer das Kernkraftwerk Goesgen-Daeniken. Zusammenfassung, Ergebnisse und Bewertung

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-15

    plant safety and radiation protection are taken into account. Even if the plant manager considers the guarantee of plant safety as his duty, an overall investigation by the authorities makes sense because it also looks into rare accident scenarios for which there are, of course, no actual working experience and which can only be considered within the framework of extended plant examinations. The PSRs on the Swiss nuclear power plants therefore complement the continuous control activities of the HSK; they are carried out about every 10 years. For KKG the PSR process was initiated by a letter from the HSK in February 1994. The areas to be considered were: a) examination of design and fulfilment of technical safety systems and comparison with the actual state-of-the-art of science and technology; b) evaluation of operational experience; c) review of the technical precautions against severe accidents including the preparation of emergency measures; d) review of the emergency organisation; e) examination of the plant protection against radioactivity; f) future dismantling at the end of operational life and disposal of the radioactive wastes; g) evaluation of accident analyses and of the KKG probabilistic safety analysis; h) review of plant organisation and plant management. The examination confirmed that, at KKG, there are very many technical safety precautions. KKG operational experience is good, the results show a high degree of operational availability and a very low number of incidental shut-downs. In international comparison the collective doses of the staff are low and the release of radioactive materials to the environment is negligible; on this account KKG is one of the world's best plants operating pressurised water reactors. Up to now the examinations have not brought any ageing deterioration to light concerning the status of safety-relevant components or ducts

  5. Demonstration of a Safety Analysis on a Complex System

    Science.gov (United States)

    Leveson, Nancy; Alfaro, Liliana; Alvarado, Christine; Brown, Molly; Hunt, Earl B.; Jaffe, Matt; Joslyn, Susan; Pinnell, Denise; Reese, Jon; Samarziya, Jeffrey; Sandys, Sean; Shaw, Alan; Zabinsky, Zelda

    1997-01-01

    For the past 17 years, Professor Leveson and her graduate students have been developing a theoretical foundation for safety in complex systems and building a methodology upon that foundation. The methodology includes special management structures and procedures, system hazard analyses, software hazard analysis, requirements modeling and analysis for completeness and safety, special software design techniques including the design of human-machine interaction, verification, operational feedback, and change analysis. The Safeware methodology is based on system safety techniques that are extended to deal with software and human error. Automation is used to enhance our ability to cope with complex systems. Identification, classification, and evaluation of hazards is done using modeling and analysis. To be effective, the models and analysis tools must consider the hardware, software, and human components in these systems. They also need to include a variety of analysis techniques and orthogonal approaches: There exists no single safety analysis or evaluation technique that can handle all aspects of complex systems. Applying only one or two may make us feel satisfied, but will produce limited results. We report here on a demonstration, performed as part of a contract with NASA Langley Research Center, of the Safeware methodology on the Center-TRACON Automation System (CTAS) portion of the air traffic control (ATC) system and procedures currently employed at the Dallas/Fort Worth (DFW) TRACON (Terminal Radar Approach CONtrol). CTAS is an automated system to assist controllers in handling arrival traffic in the DFW area. Safety is a system property, not a component property, so our safety analysis considers the entire system and not simply the automated components. Because safety analysis of a complex system is an interdisciplinary effort, our team included system engineers, software engineers, human factors experts, and cognitive psychologists.

  6. 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.

  7. Research advance in safety analysis methods for high concrete dam

    Institute of Scientific and Technical Information of China (English)

    REN; QingWen; XU; LanYu; WAN; YunHui

    2007-01-01

    High tensile stresses occurred in high concrete dams and in their foundation lead to the growing importance of their safety with the increase of concrete dam height.Without any exiting specification or successful experiences of concrete dams up to 300 m at home and abroad for reference,experts feel obliged to figure out how to perform safety analysis on high concrete dam.This paper involves the main contents and mechanical features of the safety analysis on high concrete dam and shows the current state and progress of the analysis methods.For the insufficiency and problems existing in normative methods,study on modern numerical method such as finite element method must be strengthened to find out the stress control criterion which is in accordance with the methods.Two aspects of the safety analysis of high dam--local damage from material level and integral destruction from structure level--should be considered.For the local damage,we should consider the non-homogeneity of material and strengthen the research of meso-damage mechanics.While for integral destruction of the system of high dam and its foundation,a study on non-strength theory should receive enough concerns.Further,attention should be paid to the research on the failure modes and criterions of high concrete dam failure analysis and safety evaluation,and the effect of uncertainty and classification of safety should be considered too.

  8. Thermohydraulic incidents at full power (safety analysis detailed report no. 1)

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    In this paper, attention is focused on the role of plant-incident analysis during the design stage and the conclusions reached regarding safety. This class of incidents includes sequences arising from the breakdown or anomalous behaviour of components or from errors in plant operation that have repercussions on the process of the system involved and the related systems. The sequences of possible relevance to safety are those which stress the active and passive protection (containment barriers). As these stresses are below the design-basis limits, they have no consequences in terms of radioactivity release. This report illustrates in greater detail the analysis that led to this conclusion, with particular reference to reactor events that have significant consequences on the first barrier (fuel cladding). Thermohydraulic incidents at full power are examined here.

  9. 10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

    Science.gov (United States)

    2010-01-01

    ... important to safety for nuclear power plants. 50.49 Section 50.49 Energy NUCLEAR REGULATORY COMMISSION... nuclear power plants. (a) Each holder of or an applicant for an operating license issued under this part... nuclear power plant for which the certifications required under § 50.82(a)(1) or § 52.110(a)(1) of...

  10. New Methods and Tools to Perform Safety Analysis within RISMC

    Energy Technology Data Exchange (ETDEWEB)

    Diego Mandelli; Curtis Smith; Cristian Rabiti; Andrea Alfonsi; Robert Kinoshita; Joshua Cogliati

    2013-11-01

    The Risk Informed Safety Margins Characterization (RISMC) Pathway uses a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. What differentiates the RISMC approach from traditional probabilistic risk assessment (PRA) is the concept of safety margin. In PRA, a safety metric such as core damage frequency (CDF) is generally estimated using static fault-tree and event-tree models. However, it is not possible to estimate how close we are to physical safety limits (say peak clad temperature) for most accident sequences described in the PRA. In the RISMC approach, what we want to understand is not just the frequency of an event like core damage, but how close we are (or not) to this event and how we might increase our safety margin through margin management strategies in a Dynamic PRA (DPRA) fashion. This paper gives an overview of methods that are currently under development at the Idaho National Laboratory (INL) with the scope of advance the current state of the art of dynamic PRA.

  11. 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.

  12. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Chung, D.Y.

    1999-07-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45.

  13. Safety evaluation for packaging for onsite transfer of B Plant organic waste

    Energy Technology Data Exchange (ETDEWEB)

    Mercado, M.S.

    1996-10-07

    This safety evaluation for packaging authorizes the use of a 17,500-L (4,623-gal) tank manufactured by Brenner Tank, Incorporated, to transport up to 16,221 L (4,285 gal) of radioactive organic liquid waste. The waste will be transported from the organic loading pad to a storage pad. Both pads are within the B Plant complex, but approximately 4 mi apart.

  14. Comparative analysis of twelve Dothideomycete plant pathogens

    Energy Technology Data Exchange (ETDEWEB)

    Ohm, Robin; Aerts, Andrea; Salamov, Asaf; Goodwin, Stephen B.; Grigoriev, Igor

    2011-03-11

    The Dothideomycetes are one of the largest and most diverse groups of fungi. Many are plant pathogens and pose a serious threat to agricultural crops grown for biofuel, food or feed. Most Dothideomycetes have only a single host and related Dothideomycete species can have very diverse host plants. Twelve Dothideomycete genomes have currently been sequenced by the Joint Genome Institute and other sequencing centers. They can be accessed via Mycocosm which has tools for comparative analysis

  15. 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.

  16. Technology, safety, and costs of decommissioning a reference nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Jenkins, C.E.; Rhoads, R.E.

    1977-09-01

    Safety and cost information were developed for the conceptual decommissioning of a fuel reprocessing plant with characteristics similar to the Barnwell Nuclear Fuel Plant. The main process building, spent fuel receiving and storage station, liquid radioactive waste storage tank system, and a conceptual high-level waste-solidification facility were postulated to be decommissioned. The plant was conceptually decommissioned to three decommissioning states or modes; layaway, protective storage, and dismantlement. Assuming favorable work performance, the elapsed time required to perform the decommissioning work in each mode following plant shutdown was estimated to be 2.4 years for layaway, 2.7 years for protective storage, and 5.2 years for dismantlement. In addition to these times, approximately 2 years of planning and preparation are required before plant shutdown. Costs, in constant 1975 dollars, for decommissioning were estimated to be $18 million for layaway, $19 million for protective storage and $58 million for dismantlement. Maintenance and surveillance costs were estimated to be $680,000 per year after layaway and $140,000 per year after protective storage. The combination mode of protective storage followed by dismantlement deferred for 10, 30, and 100 years was estimated to cost $64 million, $67 million and $77 million, respectively, in nondiscounted total 1975 dollars. Present values of these costs give reduced costs as dismantlement is deferred. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year radiation dose commitment to the members of the public from airborne releases from normal decommissioning activities were estimated to be less than 11 man-rem.

  17. Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

    Energy Technology Data Exchange (ETDEWEB)

    Azarm, M A; Boccio, J L; Mitra, S

    1985-12-01

    This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs.

  18. Software Safety Analysis of a Flight Guidance System

    Science.gov (United States)

    Butler, Ricky W. (Technical Monitor); Tribble, Alan C.; Miller, Steven P.; Lempia, David L.

    2004-01-01

    This document summarizes the safety analysis performed on a Flight Guidance System (FGS) requirements model. In particular, the safety properties desired of the FGS model are identified and the presence of the safety properties in the model is formally verified. Chapter 1 provides an introduction to the entire project, while Chapter 2 gives a brief overview of the problem domain, the nature of accidents, model based development, and the four-variable model. Chapter 3 outlines the approach. Chapter 4 presents the results of the traditional safety analysis techniques and illustrates how the hazardous conditions associated with the system trace into specific safety properties. Chapter 5 presents the results of the formal methods analysis technique model checking that was used to verify the presence of the safety properties in the requirements model. Finally, Chapter 6 summarizes the main conclusions of the study, first and foremost that model checking is a very effective verification technique to use on discrete models with reasonable state spaces. Additional supporting details are provided in the appendices.

  19. Emerging frontier technologies for food safety analysis and risk assessment

    Institute of Scientific and Technical Information of China (English)

    DONG Yi-yang; LIU Jia-hui; WANG Sai; CHEN Qi-long; GUO Tian-yang; ZHANG Li-ya; JIN Yong; SU Hai-jia; TAN Tian-wei

    2015-01-01

    Access to security and safe food is a basic human necessity and essential for a sustainable world. To perform hi-end food safety analysis and risk assessment with state of the art technologies is of utmost importance thereof. With applications as exempliifed by microlfuidic immunoassay, aptasensor, direct analysis in real time, high resolution mass spectrometry, benchmark dose and chemical speciifc adjustment factor, this review presents frontier food safety analysis and risk assess-ment technologies, from which both food quality and public health wil beneift undoubtedly in a foreseeable future.

  20. Aging of turbine drives for safety-related pumps in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Cox, D.F. [Oak Ridge National Lab., TN (United States)

    1995-06-01

    This study was performed to examine the relationship between time-dependent degradation and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety-related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized-water reactor plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented.

  1. Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea.

  2. National nuclear power plant safety research 2007-2010. Proposal for SAFIR2010 framework plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    A country utilising nuclear energy is presumed to possess a sufficient infrastructure to cover the education and research in this field, besides the operating organisations of the plants and a regulatory body. The starting point of public nuclear safety research programmes is that they provide the necessary conditions for retaining the knowledge needed for ensuring the continuance of safe and economic use of nuclear power, for development of new know-how and for participation in international cooperation. In fact, the Finnish organisations engaged in research in this sector have been an important resource which the various ministries, the Radiation and Nuclear Safety Authority (STUK) and the power companies have had at their disposal. Ministry of Trade and Industry appointed a group to write the Framework Plan of the new programme. This report contains a proposal for the general outline of the programme, preliminarily entitled as SAFIR2010 (SAfety of Nuclear Power Plants - Finnish National Research Programme). The plan has been made for the period 2007-2010, but it is based on safety challenges identified for a longer time span as well. Olkiluoto 3, the new nuclear power plant unit under construction has also been taken into account in the plan. The safety challenges set by the existing plants and the new plant unit, as well as the ensuing research needs do, however, converge to a great extent. The research programme is strongly based on the Chapter 7a of the Finnish Nuclear Energy Act. The construction of the new power plant unit will increase the need for experts in the field in Finland. At the same time, the retirement of the existing experts is continuing. These factors together will call for more education and training, in which active research activities play a key role. This situation also makes long-term safety research face a great challenge. The Framework Plan aims to define the important research needs related to the safety challenges. The research into

  3. The material concept in German light water reactors. Contribution to plant safety economic efficiency and failure provision; Das Werkstoffkonzept in deutschen Leichtwasserreaktoren. Beitrag zur Anlagensicherheit, Wirtschaftlichkeit und Schadensvorsorge

    Energy Technology Data Exchange (ETDEWEB)

    Ilg, Ulf [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Philippsburg; Koenig, Guenter [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Neckarwestheim; Erve, Manfred [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    In the design and construction stage of nuclear power plants relevant decisions may affect the service life of a component, and thus influence safety and availability of the plant. The German ''basic safety concept'' has an important effect on the quality of the BOL (begin of life) status. Materials selection and qualification are of significant importance for the component lifetime and the profitability of the plant. Examples for the implementation of this concept are demonstrated for the steam generator tubing material Incoloy 800, the inside-plated ferritic compound tubes as control rod drive mechanism nozzle through the RPV head of BWR plants that are not susceptible for corrosion enhanced cracking that was observed for Inconel 600 tubing. A fundamental failure analysis of crack formation in Ti stabilized austenitic pipes of BWR plants found since 1993 were definitely identified as intergranular stress corrosion caused by a local sensitization of the welding process induced overheated structured in the heat affected zone. This allowed target-oriented mitigation measures. The safety culture implemented in German nuclear plants in connection with the break preclusion or integrity concept, respectively, including a continuous actualization with respect to the state-of-the art are the technical prerequisites for damage precaution and possible life time extension.

  4. Development of safety assessment of nuclear power plants using indicators; Ydinvoimalaitosten turvallisuuden arvioinnin kehittaeminen tunnuslukujen avulla

    Energy Technology Data Exchange (ETDEWEB)

    Tiippana, P.

    1997-11-01

    The study is based on an indicator system which is under development at the Radiation and Nuclear Safety Authority (STUK). The goal of this study was to define and develop both PSA-based indicators and indicators from failure statistics. As PSA-based indicators the possibility was studied to define and express the risk importance of exemptions from the Technical Specifications, failures, preventive maintenance and other disconnections of devices covered by the Technical Specifications, operating events covered by Guide YVL 1.5 and plant modifications. In this piece of research the applicability of plant specific living PSA-models used for calculation of indicators was examined. The research included both Loviisa and Olkiluoto nuclear power plants in Finland. 47 refs.

  5. SAFETY

    CERN Document Server

    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...

  6. Health-safety and environmental risk assessment of power plants using multi criteria decision making method

    Directory of Open Access Journals (Sweden)

    Jozi Ali Seyed

    2011-01-01

    Full Text Available Growing importance of environmental issues at global and regional levels including pollution of water, air etc. as well as the outcomes such as global warming and climate change has led to being considered environmental aspects as effective factors for power generation. Study ahead, aims at examination of risks resulting from activities of Yazd Combined Cycle Power Plant located in Iran. Method applied in the research is analytical hierarchy process. After identification of factors causing risk, the analytical hierarchy structure of the power plant risks were designed and weight of the criteria and sub-criteria were calculated by intensity probability product using Eigenvector Method and EXPERT CHOICE Software as well. Results indicate that in technological, health-safety, biophysical and socio economic sections of the power plant, factors influenced by the power plant activities like fire and explosion, hearing loss, quantity of groundwater, power generation are among the most important factors causing risk in the power plant. The drop in underground water levels is the most important natural consequence influenced on Yazd Combined Cycle Power Plant.

  7. Quantitative analysis of protein turnover in plants.

    Science.gov (United States)

    Nelson, Clark J; Li, Lei; Millar, A Harvey

    2014-03-01

    Proteins are constantly being synthesised and degraded as plant cells age and as plants grow, develop and adapt the proteome. Given that plants develop through a series of events from germination to fruiting and even undertake whole organ senescence, an understanding of protein turnover as a fundamental part of this process in plants is essential. Both synthesis and degradation processes are spatially separated in a cell across its compartmented structure. The majority of protein synthesis occurs in the cytosol, while synthesis of specific components occurs inside plastids and mitochondria. Degradation of proteins occurs in both the cytosol, through the action of the plant proteasome, and in organelles and lytic structures through different protease classes. Tracking the specific synthesis and degradation rate of individual proteins can be undertaken using stable isotope feeding and the ability of peptide MS to track labelled peptide fractions over time. Mathematical modelling can be used to follow the isotope signature of newly synthesised protein as it accumulates and natural abundance proteins as they are lost through degradation. Different technical and biological constraints govern the potential for the use of (13)C, (15)N, (2)H and (18)O for these experiments in complete labelling and partial labelling strategies. Future development of quantitative protein turnover analysis will involve analysis of protein populations in complexes and subcellular compartments, assessing the effect of PTMs and integrating turnover studies into wider system biology study of plants.

  8. Environmental analysis for pipeline gas demonstration plants

    Energy Technology Data Exchange (ETDEWEB)

    Stinton, L.H.

    1978-09-01

    The Department of Energy (DOE) has implemented programs for encouraging the development and commercialization of coal-related technologies, which include coal gasification demonstration-scale activities. In support of commercialization activities the Environmental Analysis for Pipeline Gas Demonstration Plants has been prepared as a reference document to be used in evaluating potential environmental and socioeconomic effects from construction and operation of site- and process-specific projects. Effluents and associated impacts are identified for six coal gasification processes at three contrasting settings. In general, impacts from construction of a high-Btu gas demonstration plant are similar to those caused by the construction of any chemical plant of similar size. The operation of a high-Btu gas demonstration plant, however, has several unique aspects that differentiate it from other chemical plants. Offsite development (surface mining) and disposal of large quantities of waste solids constitute important sources of potential impact. In addition, air emissions require monitoring for trace metals, polycyclic aromatic hydrocarbons, phenols, and other emissions. Potential biological impacts from long-term exposure to these emissions are unknown, and additional research and data analysis may be necessary to determine such effects. Possible effects of pollutants on vegetation and human populations are discussed. The occurrence of chemical contaminants in liquid effluents and the bioaccumulation of these contaminants in aquatic organisms may lead to adverse ecological impact. Socioeconomic impacts are similar to those from a chemical plant of equivalent size and are summarized and contrasted for the three surrogate sites.

  9. Error estimation in plant growth analysis

    Directory of Open Access Journals (Sweden)

    Andrzej Gregorczyk

    2014-01-01

    Full Text Available The scheme is presented for calculation of errors of dry matter values which occur during approximation of data with growth curves, determined by the analytical method (logistic function and by the numerical method (Richards function. Further formulae are shown, which describe absolute errors of growth characteristics: Growth rate (GR, Relative growth rate (RGR, Unit leaf rate (ULR and Leaf area ratio (LAR. Calculation examples concerning the growth course of oats and maize plants are given. The critical analysis of the estimation of obtained results has been done. The purposefulness of joint application of statistical methods and error calculus in plant growth analysis has been ascertained.

  10. Determination of Initial Conditions for the Safety Analysis by Random Sampling of Operating Parameters

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Hae-Yong; Park, Moon-Ghu [Sejong University, Seoul (Korea, Republic of)

    2015-05-15

    In most existing evaluation methodologies, which follow a conservative approach, the most conservative initial conditions are searched for each transient scenario through tremendous assessment for wide operating windows or limiting conditions for operation (LCO) allowed by the operating guidelines. In this procedure, a user effect could be involved and a remarkable time and human resources are consumed. In the present study, we investigated a more effective statistical method for the selection of the most conservative initial condition by the use of random sampling of operating parameters affecting the initial conditions. A method for the determination of initial conditions based on random sampling of plant design parameters is proposed. This method is expected to be applied for the selection of the most conservative initial plant conditions in the safety analysis using a conservative evaluation methodology. In the method, it is suggested that the initial conditions of reactor coolant flow rate, pressurizer level, pressurizer pressure, and SG level are adjusted by controlling the pump rated flow, setpoints of PLCS, PPCS, and FWCS, respectively. The proposed technique is expected to contribute to eliminate the human factors introduced in the conventional safety analysis procedure and also to reduce the human resources invested in the safety evaluation of nuclear power plants.

  11. Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hollnagel, E. [Linkoeping Univ. (Sweden). Dept. of Computer and Information Science; Gauthereau, V. [Linkoeping Univ. (Sweden). Dept. of Industrial Engineering

    2001-06-01

    interviews were conducted with technical staff at most of the Swedish NPPs. It focused on which solutions the various NPPs had developed to cope with the problem, and which steps had been taken specifically to improve the efficiency of ORV. It was soon found that ORV could not be separated from the rest of the work done in a NPP during outages since many of the proposed solutions have a broad scope. An analysis of the nine Swedish ORV cases had found weaknesses in four main areas: administration processes, management, human performance, and control room layout. Relative to these, the Swedish NPPs have implemented several technical and organisational solutions. Among the former are an overall re-qualification scheme, blocked safety functions, computerised operational position control, and central indications in the control room. Most of the technical solutions have been part of the design of the newer plants, since to implement them in older plants requires essential changes both in the station and in the control room. The organisational solutions comprised operational readiness plans, systematic ways of working, new instructions, co-ordinated testing, and the use of redundant or independent controls. Special emphasis was put on how the NPPs planned their outages, how the plans were implemented, and how deviations were handled. Issues related to learning from experience were also investigated. It was found that although all the NPPs approached the ORV issues in a serious and efficient manner, the solutions could be different corresponding to the characteristics of the organisation. Finally a number of questions, which still need answers, were identified. One is how new procedures or new barriers are accepted and assimilated into the safety culture. A second concerns the demarcation of systems for which ORV is required, i.e., the boundary between safety and non-safety systems. A third is how complex technical solutions influence the operators' work. Finally, it is

  12. Safety analysis of the existing 850 Firing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Odell, B.N.

    1986-06-05

    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.

  13. Safety analysis of the existing 851 Firing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Odell, B.N.

    1986-06-05

    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.

  14. Development of the Verification and Validation Matrix for Safety Analysis Code SPACE

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yo Han; Ha, Sang Jun; Yang, Chang Keun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Korea Electric Power Research Institute (KEPRI) has been developed the safety analysis code, called as SPACE (Safety and Performance Analysis CodE for Nuclear Power Plant), for typical pressurized water reactors (PWR). Current safety analysis codes were conducted from foreign vendors, such as Westinghouse Electric Corp., ABB Combustion Engineering Inc., Kraftwerk Union, etc. Considering the conservatism and inflexibility of the foreign code systems, it is difficult to expand the application areas and analysis scopes. To overcome the mentioned problems KEPRI has launched the project to develop the native safety analysis code with Korea Power Engineering Co.(KOPEC), Korea Atomic Energy Research Inst.(KAERI), Korea Nuclear Fuel(KNF), and Korea Hydro and Nuclear Power Co.(KHNP) under the funding of Ministry of Knowledge Economy (MKE). As a result of the project, the demo-version of SPACE has been released in July 2009. As an advance preparation of the next step, KEPRI and colleagues have developed the verification and validation (V and V) matrix for SPACE. To develop the matrix, the preceding studies and experiments were reviewed. After mature consideration, the V and V matrix has been developed and the experiment plans were designed for the next step to compensate the lack of data.

  15. System safety analysis of an autonomous mobile robot

    Energy Technology Data Exchange (ETDEWEB)

    Bartos, R.J.

    1994-08-01

    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. Lithium-thionyl chloride cell system safety hazard analysis

    Science.gov (United States)

    Dampier, F. W.

    1985-03-01

    This system safety analysis for the lithium thionyl chloride cell is a critical review of the technical literature pertaining to cell safety and draws conclusions and makes recommendations based on this data. The thermodynamics and kinetics of the electrochemical reactions occurring during discharge are discussed with particular attention given to unstable SOCl2 reduction intermediates. Potentially hazardous reactions between the various cell components and discharge products or impurities that could occur during electrical or thermal abuse are described and the most hazardous conditions and reactions identified. Design factors influencing the safety of Li/SOCl2 cells, shipping and disposal methods and the toxicity of Li/SOCl2 battery components are additional safety issues that are also addressed.

  17. 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...

  18. 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.

  19. Safety estimation of structural systems via interval analysis

    Institute of Scientific and Technical Information of China (English)

    Wang Xiaojun; Wang Lei; Qiu Zhiping

    2013-01-01

    Considering that the uncertain information has serious influences on the safety of structural systems and is always limited,it is reasonable that the uncertainties are generally described as interval sets.Based on the non-probabilistic set-theoretic theory,which is applied to measuring the safety of structural components and further combined with the branch-and-bound method for the probabilistic reliability analysis of structural systems,the non-probabilistic branch-and-bound method for determining the dominant failure modes of an uncertain structural system is given.Meanwhile,a new system safety measuring index obtained by the non-probabilistic set-theoretic model is investigated.Moreover,the compatibility between the classical probabilistic model as well as the proposed interval-set model will be discussed to verify the physical meaning of the safety measure in this paper.Some numerical examples are utilized to illustrate the validity and feasibility of the developed method.

  20. Safety and nutritional assessment of GM plants and derived food and feed: the role of animal feeding trials.

    Science.gov (United States)

    2008-03-01

    In this report the various elements of the safety and nutritional assessment procedure for genetically modified (GM) plant derived food and feed are discussed, in particular the potential and limitations of animal feeding trials for the safety and nutritional testing of whole GM food and feed. The general principles for the risk assessment of GM plants and derived food and feed are followed, as described in the EFSA guidance document of the EFSA Scientific Panel on Genetically Modified Organisms. In Section 1 the mandate, scope and general principles for risk assessment of GM plant derived food and feed are discussed. Products under consideration are food and feed derived from GM plants, such as maize, soybeans, oilseed rape and cotton, modified through the introduction of one or more genes coding for agronomic input traits like herbicide tolerance and/or insect resistance. Furthermore GM plant derived food and feed, which have been obtained through extensive genetic modifications targeted at specific alterations of metabolic pathways leading to improved nutritional and/or health characteristics, such as rice containing beta-carotene, soybeans with enhanced oleic acid content, or tomato with increased concentration of flavonoids, are considered. The safety assessment of GM plants and derived food and feed follows a comparative approach, i.e. the food and feed are compared with their non-GM counterparts in order to identify intended and unintended (unexpected) differences which subsequently are assessed with respect to their potential impact on the environment, safety for humans and animals, and nutritional quality. Key elements of the assessment procedure are the molecular, compositional, phenotypic and agronomic analysis in order to identify similarities and differences between the GM plant and its near isogenic counterpart. The safety assessment is focussed on (i) the presence and characteristics of newly expressed proteins and other new constituents and possible

  1. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  2. Factor analysis of nursing students' perception of patient safety education.

    Science.gov (United States)

    Mansour, Mansour

    2015-01-01

    The aim of this study is to investigate the factor structure of the Health Care Professionals Patient Safety Assessment Curriculum Survey (HPPSACS) when completed by a group of nursing students from one University in the UK. The quality, content and delivery of nursing education can have a significant impact on the future students' safety behaviours in clinical settings. The Health Care Professionals Patient Safety Assessment Curriculum Survey HPPSACS has been developed in the US to establish undergraduate nursing students' perceived awareness, skills, and attitudes toward patient safety education. The instrument has not been reported to be used elsewhere; therefore, some psychometric properties remain untested. Pre-registration nursing students (n=272) from three campuses of a university in East of England completed the HPPSACS in 2012. Principal component analysis was conducted to explore the factors emerging from the students' responses. 222 students (82%) returned the questionnaires. Constraining data to a 4-factor solution explained 52% of the variance. Factors identified were: "Willingness to disclose errors", "Recognition and management of medical errors", "The Perceived interprofessional context of patient safety" and "The perceived support and understanding for improving patient safety". The overall Cronbach's alpha was 0.64, indicating moderate internal consistency of the instrument. Some demographical and descriptive questions on the HPPSACS instrument were modified to accommodate the participants' educational context. However, all items in the HPPSACS which were included in the factor analysis remain identical to the original tool. The study offers empirical findings of how patient safety education is contextualised in the undergraduate, pre-registration nursing curriculum. Further research is required to refine and improve the overall reliability of the Health Care Professionals Patient Safety Assessment Curriculum Survey (HPPSACS' instrument

  3. Safety-Related Contractor Activities at Nuclear Power Plants. New Challenges for Regulatory Oversight

    Energy Technology Data Exchange (ETDEWEB)

    Chockie, Alan [Chockie Group International, Inc., Seattle, WA (United States)

    2005-09-15

    The use of contractors has been an integral and important part of the design, construction, operation, and maintenance of nuclear power plants. To ensure the safe and efficient completion of contracted tasks, each nuclear plant licensee has developed and refined formal contract management processes to meet their specific needs and plant requirements. Although these contract management processes have proven to be effective tools for the procurement of support and components tailored to the needs of nuclear power plants, contractor-related incidents and accidents have revealed some serious weaknesses with the implementation of these processes. Identifying and addressing implementation problems are becoming more complicated due to organizational and personnel changes affecting the nuclear power industry. The ability of regulators and licensees to effectively monitor and manage the safety-related performance of contractors will likely be affected by forthcoming organization and personnel changes due to: the aging of the workforce; the decline of the nuclear industry; and the deregulation of nuclear power. The objective of this report is to provide a review of current and potential future challenges facing safety-related contractor activities at nuclear power plants. The purpose is to assist SKI in establishing a strategy for the proactive oversight of contractor safety-related activities at Swedish nuclear power plants and facilities. The nature and role of contractors at nuclear plants is briefly reviewed in the first section of the report. The second section describes the essential elements of the contract management process. Although organizations have had decades of experience with the a contract management process, there remain a number of common implantation weaknesses that have lead to serious contractor-related incidents and accidents. These implementation weaknesses are summarized in the third section. The fourth section of the report highlights the

  4. Plant stress analysis technology deployment

    Energy Technology Data Exchange (ETDEWEB)

    Ebadian, M.A.

    1998-01-01

    Monitoring vegetation is an active area of laser-induced fluorescence imaging (LIFI) research. The Hemispheric Center for Environmental Technology (HCET) at Florida International University (FIU) is assisting in the transfer of the LIFI technology to the agricultural private sector through a market survey. The market survey will help identify the key eco-agricultural issues of the nations that could benefit from the use of sensor technologies developed by the Office of Science and Technology (OST). The principal region of interest is the Western Hemisphere, particularly, the rapidly growing countries of Latin America and the Caribbean. The analysis of needs will assure that the focus of present and future research will center on economically important issues facing both hemispheres. The application of the technology will be useful to the agriculture industry for airborne crop analysis as well as in the detection and characterization of contaminated sites by monitoring vegetation. LIFI airborne and close-proximity systems will be evaluated as stand-alone technologies and additions to existing sensor technologies that have been used to monitor crops in the field and in storage.

  5. Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 1. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C. E.; Murphy, E. S.; Schneider, K J

    1979-01-01

    Detailed technology, safety and cost information are presented for the conceptual decommissioning of a reference small mixed oxide fuel fabrication plant. Alternate methods of decommissioning are described including immediate dismantlement, safe storage for a period of time followed by dismantlement and entombment. Safety analyses, both occupational and public, and cost evaluations were conducted for each mode.

  6. Safety and nutritional assessment of GM plants and derived food and feed: The role of animal feeding trials

    NARCIS (Netherlands)

    Haver, van E.; Alink, G.M.; Cockburn, A.; Kuiper, H.A.; Peijnenburg, A.A.C.M.

    2008-01-01

    In this report the various elements of the safety and nutritional assessment procedure for genetically modified (GM) plant derived food and feed are discussed, in particular the potential and limitations of animal feeding trials for the safety and nutritional testing of whole GM food and feed. The g

  7. Safety and nutritional assessment of GM plants and derived food and feed: The role of animal feeding trials

    NARCIS (Netherlands)

    Haver, van E.; Alink, G.M.; Cockburn, A.; Kuiper, H.A.; Peijnenburg, A.A.C.M.

    2008-01-01

    In this report the various elements of the safety and nutritional assessment procedure for genetically modified (GM) plant derived food and feed are discussed, in particular the potential and limitations of animal feeding trials for the safety and nutritional testing of whole GM food and feed. The

  8. 77 FR 50720 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2012-08-22

    ... COMMISSION Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants... regulatory guide (DG), DG-1207, ``Test Documentation for Digital Computer Software used in Safety Systems of... revision endorses, with clarifications, the enhanced consensus practices for test documentation...

  9. Value of Information Analysis in Structural Safety

    DEFF Research Database (Denmark)

    Konakli, Katerina; Faber, Michael Havbro

    2014-01-01

    of structural systems. In this context, experiments may refer to inspections or techniques of structural health monitoring. The Value of Information concept provides a powerful tool for determining whether the experimental cost is justified by the expected benefit and for identifying the optimal among different......Pre-posterior analysis can be used to assess the potential of an experiment to enhance decision making by providing information on parameters characterized by uncertainty. The present paper describes a framework for pre-posterior analysis for support of decisions related to maintenance...... and quality of information and the probabilistic dependencies between components of a system....

  10. One safety critical indicators model for regulatory actions on nuclear power plants based on a level 1 PSA; Um modelo de indicadores criticos de seguranca para acoes regulatorias em usinas nucleares baseado em uma APS nivel 1

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Jefferson Borges

    2006-03-15

    This study presents a general methodology to the establishment, selection and use of safety indicators for a two loop PWR plant, as Angra 1. The study performed identifies areas considered critical for the plant operational safety. For each of these areas, strategic sub-areas are defined. For each strategic sub-area, specific safety indicators are defined. These proposed Safety Indicators are based on the contribution to risk considering a quantitative risk analysis. For each safety indicator, a goal, a bounded interval and proper bases are developed, to allow for a clear and comprehensive individual behavior evaluation. Additionally, an integrated evaluation of the indicators, using expert systems, was done to obtain an overview of the plant general safety. This methodology can be used for identifying situations where the plant safety is challenged, by giving a general overview of the plant operational condition. Additionally, this study can also identify eventual room for improvements by generating suggestions and recommendations, as a complement for regulatory actions and inspections, focusing resources on eventual existing weaknesses, in order to increase or maintain a high pattern of operational safety. (author)

  11. Safety

    CERN Multimedia

    2003-01-01

    Please note that the safety codes A9, A10 AND A11 (ex annexes of SAPOCO/42) entitled respectively "Safety responsibilities in the divisions" "The safety policy committee (SAPOCO) and safety officers' committees" and "Administrative procedure following a serious accident or incident" are available on the web at the following URLs: Code A9: http://edms.cern.ch/document/337016/LAST_RELEASED Code A10: http://edms.cern.ch/document/337019/LAST_RELEASED Code A11: http://edms.cern.ch/document/337026/LAST_RELEASED Paper copies can also be obtained from the TIS divisional secretariat, e-mail: tis.secretariat@cern.ch. TIS Secretariat

  12. Risk-Based Explosive Safety Analysis

    Science.gov (United States)

    2016-11-30

    other provision of law, no person shall be subject to any penalty for failing to comply with a collection of information if it does not display a...based analysis of scenario 2 would likely determine that the hazard of death or injury to any single person is low due to the separation distance

  13. Comparison of Hazard Analysis Requirements for Instrumentation and Control System of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo [KAERI, Daejeon (Korea, Republic of); Yoo, Jun Beom [Konkuk University, Seoul (Korea, Republic of)

    2014-08-15

    A hazard, in general, is defined as 'potential for harm.' In this paper, the scope of 'harm' is limited to the loss of a safety function in a Nuclear Power Plant (NPP). The Hazard Analysis (HA) of an Instrumentation and Control (I and C) systems is to identify the relationship from the logical faults, error, and failure of I and C systems to the physical harm of the nuclear power plant, and also to find the impact of the external hazard, e.g., tsunami, of the nuclear power plant to the I and C systems. This paper includes the survey of the existing hazard analysis requirements in the nuclear industries. The purpose of the paper is to compare the HA requirements in various international standards in unclear domain, specifically the safety requirements and guidance for the instrumentation and control system for the nuclear power plant from IAEA, IEC, IEEE, and NRC.

  14. 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.

  15. The safety assessment of food ingredients derived from plant cell, tissue and organ cultures: a review.

    Science.gov (United States)

    Murthy, Hosakatte Niranjana; Georgiev, Milen I; Park, So-Young; Dandin, Vijayalaxmi S; Paek, Kee-Yoeup

    2015-06-01

    Plant cell, tissue and organ cultures (PCTOC) have become an increasingly attractive alternative for the production of various high molecular weight molecules which are used as flavourings, fragrances, colouring agents and food additives. Although PCTOC products are cultivated in vitro in a contamination free environment, the raw material produced from PCTOC may contain many components apart from the target compound. In some cases, PCTOC raw materials may also carry toxins, which may be naturally occurring or accumulated during the culture process. Assessment of the safety of PCTOC products is, therefore, a priority of the biotech industries involved in their production. The safety assessment involves the evaluation of starting material, production process and the end product. Before commercialisation, PCTOC products should be evaluated for their chemical and biological properties, as well as for their toxicity. In this review, measures and general criteria for biosafety evaluation of PCTOC products are addressed and thoroughly discussed.

  16. Resolving plant operational issues related to pressurizer safety relief valve piping

    Energy Technology Data Exchange (ETDEWEB)

    Bain, R.A. [Stone and Webster Engineering Corp., Boston, MA (United States). Mechanical Engineering Div.; Testa, M.F. [Duquesne Light Co., Shippingport, PA (United States). Mechanical Engineering Dept.

    1995-11-01

    Pressurizer safety and relief valve (PSARV) piping has many technical issues related to the qualification and operation of the system that have been addressed at Beaver Valley Unit 2. The PSARV piping is part of a system that must meet Code requirements while being subjected to very significant fluid transient loadings. Valve components include safety valves, power operated relief valves (PORVs), and motor operated block valves. Fluid slugs upstream of these valves can be steam or can be hot or cold water, resulting in a significant variance in possible slug densities. Problems with design options and hardware installed to decrease slug density such as heat tracing, and the susceptibility of the PORVs to leak are issues that affect plant operation, efficiency and cost effectiveness.

  17. Valuation of road safety effects in cost-benefit analysis.

    Science.gov (United States)

    Wijnen, Wim; Wesemann, Paul; de Blaeij, Arianne

    2009-11-01

    Cost-benefit analysis is a common method for evaluating the social economic impact of transport projects, and in many of these projects the saving of human lives is an issue. This implies, within the framework of cost-benefit analysis, that a monetary value should be attached to saving human lives. This paper discusses the 'Value of a Statistical Life' (VoSL), a concept that is often used for monetising safety effects, in the context of road safety. Firstly, the concept of 'willingness to pay' for road safety and its relation to the VoSL are explained. The VoSL approach will be compared to other approaches to monetise safety effects, in particular the human capital approach and 'quality adjusted life years'. Secondly, methods to estimate the VoSL and their applicability to road safety will be discussed. Thirdly, the paper reviews the VoSL estimates that have been found in scientific research and compares them with the values that are used in policy evaluations. Finally, a VoSL study in the Netherlands will be presented as a case study, and its applicability in policy evaluation will be illustrated.

  18. Model Based Safety Analysis with smartIflow †

    Directory of Open Access Journals (Sweden)

    Philipp Hönig

    2017-01-01

    Full Text Available Verification of safety requirements is one important task during the development of safety critical systems. The increasing complexity of systems makes manual analysis almost impossible. This paper introduces a new methodology for formal verification of technical systems with smartIflow (State Machines for Automation of Reliability-related Tasks using Information FLOWs. smartIflow is a new modeling language that has been especially designed for the purpose of automating the safety analysis process in early product life cycle stages. It builds up on experience with existing approaches. As is common practice in current approaches, components are modeled as finite state machines. However, new concepts are introduced to describe component interactions. Events play a major role for internal interactions between components as well as for external (user interactions. Our approach to the verification of formally specified safety requirements is a two-step method. First, an exhaustive simulation creates knowledge about a great variety of possible behaviors of the system, especially including reactions on suddenly occurring (possibly intermittent faults. In the second step, safety requirements specified in CTL (Computation Tree Logic are verified using model checking techniques, and counterexamples are generated if these are not satisfied. The practical applicability of this approach is demonstrated based on a Java implementation using a simple Two-Tank-Pump-Consumer system.

  19. 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

  20. General aviation air traffic pattern safety analysis

    Science.gov (United States)

    Parker, L. C.

    1973-01-01

    A concept is described for evaluating the general aviation mid-air collision hazard in uncontrolled terminal airspace. Three-dimensional traffic pattern measurements were conducted at uncontrolled and controlled airports. Computer programs for data reduction, storage retrieval and statistical analysis have been developed. Initial general aviation air traffic pattern characteristics are presented. These preliminary results indicate that patterns are highly divergent from the expected standard pattern, and that pattern procedures observed can affect the ability of pilots to see and avoid each other.

  1. A quantitative risk assessment tool for the external safety of industrial plants with a dust explosion hazard

    NARCIS (Netherlands)

    Voort, M.M. van der; Klein, A.J.J.; Maaijer, M. de; Berg, A.C. van den; Deursen, J.R. van; Versloot, N.H.A.

    2007-01-01

    A quantitative risk assessment (QRA) tool has been developed by TNO for the external safety of industrial plants with a dust explosion hazard. As a first step an industrial plant is divided into groups of modules, defined by their size, shape, and constructional properties. Then the relevant explosi

  2. Analysis on Pollution Factors in Asparagus Production and Research on Safety Production Technology

    Institute of Scientific and Technical Information of China (English)

    Liping; MA; Bianqing; HAO; Xiongwu; QIAO

    2013-01-01

    Based on the analysis on the infection degree,infection law and influencing factors of the main diseases on asparagus and the analysis on the pollution factors in asparagus production such as blind pesticide use,atmospheric pollution and acid rain,the pollution of soil and fertilizer,this article proposes asparagus safety production technologies which include the selection of disease-resistant variety and suitable planting field,scientific and reasonable disease control,balanced fertilization,rational irrigation,making a good job of field management, etc.,to reduce pathogenic factors.

  3. Safety Implementation of Hydrogen Igniters and Recombiners for Nuclear Power Plant Severe Accident Management

    Institute of Scientific and Technical Information of China (English)

    XIAO Jianjun; ZHOU Zhiwei; JING Xingqing

    2006-01-01

    Hydrogen combustion in a nuclear power plant containment building may threaten the integrity of the containment. Hydrogen recombiners and igniters are two methods to reduce hydrogen levels in containment buildings during severe accidents. The purpose of this paper is to evaluate the safety implementation of hydrogen igniters and recombiners. This paper analyzes the risk of deliberate hydrogen ignition and investigates three mitigation measures using igniters only, hydrogen recombiners only or a combination of recombiners and igniters. The results indicate that steam can effectively control the hydrogen flame acceleration and the deflagration-to-detonation transition.

  4. Safety Analysis for Packaging Steel Banded Wooden Shipping Containers

    Energy Technology Data Exchange (ETDEWEB)

    FERRELL, P.C.

    2000-12-05

    This safety analysis report for packaging describes the steel banded wooden shipping containers, which are certified as Type AF packagings. The authorized payload for these containers is unirradiated, slightly enriched, uranium ingots, billets, extrusions, and scrap materials. The amount of uranium in the containers will not exceed the LSA-II material requirements as defined in 49 CFR 173.403.

  5. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  6. Seismic margin analysis technique for nuclear power plant structures

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jeong Moon; Choi, In Kil

    2001-04-01

    In general, the Seismic Probabilistic Risk Assessment (SPRA) and the Seismic Margin Assessment(SAM) are used for the evaluation of realistic seismic capacity of nuclear power plant structures. Seismic PRA is a systematic process to evaluate the seismic safety of nuclear power plant. In our country, SPRA has been used to perform the probabilistic safety assessment for the earthquake event. SMA is a simple and cost effective manner to quantify the seismic margin of individual structural elements. This study was performed to improve the reliability of SMA results and to confirm the assessment procedure. To achieve this goal, review for the current status of the techniques and procedures was performed. Two methodologies, CDFM (Conservative Deterministic Failure Margin) sponsored by NRC and FA (Fragility Analysis) sponsored by EPRI, were developed for the seismic margin review of NPP structures. FA method was originally developed for Seismic PRA. CDFM approach is more amenable to use by experienced design engineers including utility staff design engineers. In this study, detailed review on the procedures of CDFM and FA methodology was performed.

  7. Guidelines for nuclear-power-plant safety-issue-prioritization information development

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, W.B.; Gallucci, R.H.V.; Heaberlin, S.W.; Bickford, W.E.; Konzek, G.J.; Strenge, D.L.; Smith, R.I.; Weakley, S.A.

    1983-02-01

    Pacific Northwest Laboratory has developed a methodology, with examples, to calculate - to an approximation serviceable for prioritization purposes - the risk, dose and cost impacts of implementing resolutions to reactor safety issues. This report is an applications guide to issue-specific calculations. A description of the approach, mathematical models, worksheets and step-by-step examples are provided. Analysis using this method is intended to provide comparable results for many issues at a cost of two staff-weeks per issue. Results will be used by the NRC to support decisions related to issue priorities in allocation of resources to complete safety issue resolutions.

  8. 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.

  9. Development of Behavioral Indicators of Competences for Safety Culture of Nuclear Power Plants: A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Kwangsu; Kim, Sa Kil; Oh, Yeon Ju; Shin, Youmin; Lee, Yong-Hee; Jang, Tong Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The term of safety competency in nuclear field was presented in the OECD/NEA workshop held in 1999. A model of the safety culture competencies in nuclear power plants was developed by KAERI (Korea Atomic Energy Research Institute). In general, a competency (competence) is defined as 'cluster of employee's attribute, knowledge, skill, ability or other characteristic that contributes to successful job performance'. We also defined safety culture competency as 'cluster of various internal characteristics (e.g., knowledge, skill, ability, motive, attitude and etc.) of employee that contribute to perform job safely and shape a healthy and strong safety culture.' By this definition, the safety culture competency is the broader construct including job competency. An employee having high level of safety culture competency shows extra discretionary effort to improve safety of peer, team and organization in addition to the individual's successful and safe job accomplishment. The behavioral indicators for each of the competencies are focal points of conversations on progress and are monitored continuously by self-assessment and managers or supervisors' intervention. Deficiencies in any of these indicators can point to coaching, training or other learning opportunities that employees may be required in order to improve. The purpose of this study was to derive a model of safety competencies for improving safety culture of NPPs and develop a set of behavioral indicators of each competency. In addition, the method of measuring behavioral indicators was suggested. For the application of developed safety culture competences and behavioral indicators, the most suitable measuring method for behavioral indicators must be developed. In the case of behavioral observations, behavioral dimensions (frequency, persistence and latency), observation possibility, occurrence basis of behavior (daily job performance, situational dependent) are considered to

  10. Probabilistic safety goals for nuclear power plants; Phases 2-4. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bengtsson, L.; Knochenhauer, M. (Scandpower AB (Sweden)); Holmberg, J.-E.; Rossi, J. (VTT Technical Research Centre of Finland (Finland))

    2011-05-15

    Safety goals are defined in different ways in different countries and also used differently. Many countries are presently developing them in connection to the transfer to risk-informed regulation of both operating nuclear power plants (NPP) and new designs. However, it is far from self-evident how probabilistic safety criteria should be defined and used. On one hand, experience indicates that safety goals are valuable tools for the interpretation of results from a probabilistic safety assessment (PSA), and they tend to enhance the realism of a risk assessment. On the other hand, strict use of probabilistic criteria is usually avoided. A major problem is the large number of different uncertainties in a PSA model, which makes it difficult to demonstrate the compliance with a probabilistic criterion. Further, it has been seen that PSA results can change a lot over time due to scope extensions, revised operating experience data, method development, changes in system requirements, or increases of level of detail, mostly leading to an increase of the frequency of the calculated risk. This can cause a problem of consistency in the judgments. This report presents the results from the second, third and fourth phases of the project (2007-2009), which have dealt with providing guidance related to the resolution of some specific problems, such as the problem of consistency in judgement, comparability of safety goals used in different industries, the relationship between criteria on different levels, and relations between criteria for level 2 and 3 PSA. In parallel, additional context information has been provided. This was achieved by extending the international overview by contributing to and benefiting from a survey on PSA safety criteria which was initiated in 2006 within the OECD/NEA Working Group Risk. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by

  11. Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H. K.

    1981-10-01

    Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0

  12. Multivariate time series analysis of SafetyNet data. SafetyNet, Building the European Road Safety Observatory, Workpackage 7, Deliverable 7.7.

    NARCIS (Netherlands)

    Commandeur, J.J.F. Bijleveld, F.D. & Bergel, R.

    2009-01-01

    This deliverable provides an application of theories and methods documented in Deliverables 7.4 and 7.5 of work package 7 of the SafetyNet project. In this deliverable, use of select analysis techniques is demonstrated through real world road safety analysis problems, using aggregate data which may

  13. Application of Solar Chimney Concept to Solve Potential Safety Issues of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Khasawneh, Khalid; PARK, Youn Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    In this paper two main events and their causes have been investigated and a potential alternative supporting system will be provided. The first event to be addressed is the Station Blackout (SBO) caused by the inherent unreliability of the Emergency Diesel Generators (EDGs) and Alternative AC (AAC) power sources. Different parameters affect The EDG unreliability; for instance, mechanical, operational, maintenance and surveillance. Those parameters will be analyzed and linked to plant safety and Core Damage Frequency (CDF). Also the AACs, the SBO diesel generators, will be studied and their operational requirements similarity with the EDGs will be discussed. The second event to be addressed is the Loss of Ultimate Heat Sink (LUHS) caused by the degradation of heat exchange effectiveness, that is, the poor heat transfer to the Ultimate Heat Sink (UHS). Different causes to such case were observed; intake lines blockages due to ice and foreign biological matters formation and oil spill near the heat sink causing the oil leakage to the heat exchangers tubes. The later cause, oil spill, has been given a special attention here due its potential effects for different nuclear power plants (NPPs) around the world; for example, Finland and the United Arab Emirates (UAE). For the Finnish case, the Finnish nuclear regulator (STUK) took already countermeasures for such scenario by introducing alternative heat sink, cooling towers, for the primary used heat sink, sea water, for one of its nuclear power plants. The abundance of the solar irradiation in the UAE region provides a perfect condition for the implementation of solar power applications. Utilizing this unique characteristic of that region may provide promising alternative and diverse options for solving potential safety related issues of their NPPs. The Solar Chimney Power Plant (SCPP) could be employed to serve as a supporting system to provide emergency power, in the case of SBO, and emergency cooling, in the case of

  14. 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.

  15. 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.

  16. Application of causality diagram in system safety analysis

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Causality Diagram (CD) is a new graphical knowledge representation based on probability theory. The application of this methodology in the safety analysis of the gas explosion in collieries was discussed in this paper, and the Minimal Cut Set, the Minimal Path Set and the Importance were introduced to develop the methodology. These concepts are employed to analyze the influence each event has on the top event ( the gas explosion, so as to find out about the defects of the system and accordingly help to work out the emphasis of the precautionary work and some preventive measures as well. The results of the safety analysis are in accordance with the practical requirements; therefore the preventive measures are certain to work effectively. In brief, according to the research CD is so effective in the safety analysis and the safety assessment that it can be a qualitative and quantitative method to predict the accident as well as offer some effective measures for the investigation, the prevention and the control of the accident.

  17. 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)

  18. Exergoeconomic analysis of a cogeneration plant

    Energy Technology Data Exchange (ETDEWEB)

    Mert, Mehmet Selcuk [Yalova University, Energy Systems Engineering Department (Turkey)], email: msmert@yalova.edu.tr; Dilmac, Omer Faruk; Ozkan, Semra; Bolat, Esen [Yildiz Technical University Chemical Engineering Department (Turkey)], email: omerfarukdl@yahoo.com, email: ozkans@yildiz.edu.tr, email: ebolat@yildiz.edu.tr

    2011-07-01

    Cogeneration, or combined heat and power, is the use of a heat engine or a power station to simultaneously generate both electricity and useful heat generally in the form of steam or hot water. By capturing the excess heat, it uses thermal energy that would be wasted in a conventional power plant, provides high energy conversion efficiency, and also reduces emissions. Exergy analysis has been found to be an effective tool for predicting the efficiencies of system components and the total thermodynamic effectiveness of the system, as well as for highlighting possible improvements. By using both exergy concepts and economics, an exergoeconomic analysis is possible which helps designers to find ways to improve the performance of a system in a cost effective way. This study presents the exergoeconomic analysis of a cogeneration plant at Erdemir, Turkey. The mass, energy and exergy balances were done, and the exergoeconomic analysis of the plant, with a 39.5 MW electricity total net power and 20.83 kg/s steam production capacity, was also done.

  19. Exergoeconomic analysis of a nuclear power plant

    Science.gov (United States)

    Moreno, Roman Miguel

    Exergoeconomic analysis of a nuclear power plant is a focus of this dissertation. Specifically, the performance of the Palo Verde Nuclear Power Plant in Arizona is examined. The analysis combines thermodynamic second law exergy analysis with economics in order to assign costs to the loss and destruction of exergy. This work was done entirely with an interacting spreadsheets notebook. The procedures are to first determine conventional energy flow, where the thermodynamic stream state points are calculated automatically. Exergy flow is then evaluated along with destruction and losses. The capital cost and fixed investment rate used for the economics do not apply specifically to the Palo Verde Plant. Exergy costing is done next involving the solution of about 90 equations by matrix inversion. Finally, the analysis assigns cost to the exergy destruction and losses in each component. In this work, the cost of electricity (exergy), including capital cost, leaving the generator came to 38,400 /hr. The major exergy destruction occurs in the reactor where fission energy transfer is limited by the maxiμm permissible clad temperature. Exergy destruction costs were: reactor--18,207 hr, the low pressure turbine-2,000 /hr, the condenser--1,700 hr, the steam generator-1,200 $/hr. The inclusion of capital cost and O&M are important in new system design assessments. When investigating operational performance, however, these are sunk costs; only fuel cost needs to be considered. The application of a case study is included based on a real modification instituted at Palo Verde to reduce corrosion steam generator problems; the pressure in the steam generator was reduced from 1072 to 980 psi. Exergy destruction costs increased in the low pressure turbine and in the steam generator, but decreased in the reactor vessel and the condenser. The dissertation demonstrates the procedures and tools required for exergoeconomic analysis whether in the evaluation of a new nuclear reactor system

  20. Style, content and format guide for writing safety analysis documents. Volume 1, Safety analysis reports for DOE nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The purpose of Volume 1 of this 4-volume style guide is to furnish guidelines on writing and publishing Safety Analysis Reports (SARs) for DOE nuclear facilities at Sandia National Laboratories. The scope of Volume 1 encompasses not only the general guidelines for writing and publishing, but also the prescribed topics/appendices contents along with examples from typical SARs for DOE nuclear facilities.

  1. A computer based living probabilistic safety assessment (LPSA) method for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zubair, Muhammad, E-mail: zubairheu@gmail.com [Department of Nuclear Engineering, Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of); Department of Basic Sciences, University of Engineering and Technology, Taxila (Pakistan); Zhijian, Zhang [College of Nuclear Science and Technology, Harbin Engineering University (China); Heo, Gyunyoung [Department of Nuclear Engineering, Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of); Ahmed, Iftikhar [College of Mathematics and Statics, Chongqing University, 401331 (China); Aamir, Muhammad [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, Chongqing 400030 (China)

    2013-12-15

    Highlights: • A computer based LPSA method named, online risk monitor system (ORMS) has been proposed. • The essential features and functions of ORMS have been described. • A case study of emergency diesel generator (EDG) of Daya Bay NPP had carried out. • By using ORMS operational failure rate and demand failure probability of EDG has been calculated. - Abstract: To update PSA (probabilistic safety assessment) model this paper presents a computer based living probabilistic safety assessment (LPSA) method named as online risk monitor system (ORMS). The essential features and functions of ORMS have been described in this research. A case study of emergency diesel generator (EDG) of Daya Bay nuclear power plant (NPP) has been done; operational failure rate and demand failure probability of EDG has been calculated with the help of ORMS. The results of ORMS are well matched with data obtained from Daya Bay NPP. ORMS is capable of automatically update the online risk models and reliability parameters of equipment in time. ORMS can support in decision making process of operator and manager in nuclear power plant.

  2. Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    O, J.M.; Higgins, J.; Stephen Fleger - NRC

    2011-09-19

    The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.

  3. [Genetically modified plants and food safety. State of the art and discussion in the European Union].

    Science.gov (United States)

    Schauzu, M

    2004-09-01

    Placing genetically modified (GM) plants and derived products on the European Union's (EU) market has been regulated by a Community Directive since 1990. This directive was complemented by a regulation specific for genetically modified and other novel foods in 1997. Specific labelling requirements have been applicable for GM foods since 1998. The law requires a pre-market safety assessment for which criteria have been elaborated and continuously adapted in accordance with the state of the art by national and international bodies and organisations. Consequently, only genetically modified products that have been demonstrated to be as safe as their conventional counterparts can be commercialized. However, the poor acceptance of genetically modified foods has led to a de facto moratorium since 1998. It is based on the lack of a qualified majority of EU member states necessary for authorization to place genetically modified plants and derived foods on the market. New Community Regulations are intended to end this moratorium by providing a harmonized and transparent safety assessment, a centralised authorization procedure, extended labelling provisions and a traceability system for genetically modified organisms (GMO) and derived food and feed.

  4. Technical basis for environmental qualification of computer-based safety systems in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States); Antonescu, C.E. [Nuclear Regulatory Commission, Rockville, MD (United States)

    1997-10-01

    This paper summarizes the results of research sponsored by the US Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. This research was conducted by the Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL). ORNL investigated potential failure modes and vulnerabilities of microprocessor-based technologies to environmental stressors, including electromagnetic/radio-frequency interference, temperature, humidity, and smoke exposure. An experimental digital safety channel (EDSC) was constructed for the tests. SNL performed smoke exposure tests on digital components and circuit boards to determine failure mechanisms and the effect of different packaging techniques on smoke susceptibility. These studies are expected to provide recommendations for environmental qualification of digital safety systems by addressing the following: (1) adequacy of the present preferred test methods for qualification of digital I and C systems; (2) preferred standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging in qualification testing for equipment that is to be located in mild environments; and (5) determination of an appropriate approach to address smoke in a qualification program.

  5. Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)

    2016-09-15

    At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

  6. Status of generic actions items and safety analysis system of PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Min, Byung Joo

    2001-05-01

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

  7. Hazardous organic compounds in biogas plant end products-Soil burden and risk to food safety

    Energy Technology Data Exchange (ETDEWEB)

    Suominen, K., E-mail: kimmo.suominen@evira.fi [Finnish Food Safety Authority Evira, Risk Assessment Research Unit, Mustialankatu 3, 00790 Helsinki (Finland); Verta, M. [Finnish Environmental Institute (SYKE), Mechelininkatu 34a, P.O. Box 140, 00251 Helsinki (Finland); Marttinen, S. [MTT Agrifood Research Finland, 31600 Jokioinen (Finland)

    2014-09-01

    The end products (digestate, solid fraction of the digestate, liquid fraction of the digestate) of ten biogas production lines in Finland were analyzed for ten hazardous organic compounds or compound groups: polychlorinated dibenzo-p-dioxins and furans (PCDD/Fs), polychlorinated biphenyls (PCB(7)), polyaromatic hydrocarbons (PAH(16)), bis-(2-ethylhexyl) phthalate (DEHP), perfluorinated alkyl compounds (PFCs), linear alkylbenzene sulfonates (LASs), nonylphenols and nonylphenol ethoxylates (NP + NPEOs), polybrominated diphenyl ethers (PBDEs), hexabromocyclododecane (HBCD) and tetrabromobisphenol A (TBBPA). Biogas plant feedstocks were divided into six groups: municipal sewage sludge, municipal biowaste, fat, food industry by-products, animal manure and others (consisting of milling by-products (husk) and raw former foodstuffs of animal origin from the retail trade). There was no clear connection between the origin of the feedstocks of a plant and the concentrations of hazardous organic compounds in the digestate. For PCDD/Fs and for DEHP, the median soil burden of the compound after a single addition of digestate was similar to the annual atmospheric deposition of the compound or compound group in Finland or other Nordic countries. For PFCs, the median soil burden was somewhat lower than the atmospheric deposition in Finland or Sweden. For NP + NPEOs, the soil burden was somewhat higher than the atmospheric deposition in Denmark. The median soil burden of PBDEs was 400 to 1000 times higher than the PBDE air deposition in Finland or in Sweden. With PBDEs, PFCs and HBCD, the impact of the use of end products should be a focus of further research. Highly persistent compounds, such as PBDE- and PFC-compounds may accumulate in agricultural soil after repeated use of organic fertilizers containing these compounds. For other compounds included in this study, agricultural use of biogas plant end products is unlikely to cause risk to food safety in Finland. - Highlights:

  8. A Demonstration of Advanced Safety Analysis Tools and Methods Applied to Large Break LOCA and Fuel Analysis for PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Laboratory; Smith, Curtis Lee [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-03-01

    The U.S. Nuclear Regulatory Commission (NRC) is currently proposing a rulemaking designated as 10 CFR 50.46c to revise the loss-of-coolant accident (LOCA)/emergency core cooling system acceptance criteria to include the effects of higher burnup on fuel/cladding performance. We propose a demonstration problem of a representative four-loop PWR plant to study the impact of this new rule in the US nuclear fleet. Within the scope of evaluation for the 10 CFR 50.46c rule, aspects of safety, operations, and economics are considered in the industry application demonstration presented in this paper. An advanced safety analysis approach is used, by integrating the probabilistic element with deterministic methods for LOCA analysis, a novel approach to solving these types of multi-physics, multi-scale problems.

  9. Fire hazard analysis for Plutonium Finishing Plant complex

    Energy Technology Data Exchange (ETDEWEB)

    MCKINNIS, D.L.

    1999-02-23

    A fire hazards analysis (FHA) was performed for the Plutonium Finishing Plant (PFP) Complex at the Department of Energy (DOE) Hanford site. The scope of the FHA focuses on the nuclear facilities/structures in the Complex. The analysis was conducted in accordance with RLID 5480.7, [DOE Directive RLID 5480.7, 1/17/94] and DOE Order 5480.7A, ''Fire Protection'' [DOE Order 5480.7A, 2/17/93] and addresses each of the sixteen principle elements outlined in paragraph 9.a(3) of the Order. The elements are addressed in terms of the fire protection objectives stated in paragraph 4 of DOE 5480.7A. In addition, the FHA also complies with WHC-CM-4-41, Fire Protection Program Manual, Section 3.4 [1994] and WHC-SD-GN-FHA-30001, Rev. 0 [WHC, 1994]. Objectives of the FHA are to determine: (1) the fire hazards that expose the PFP facilities, or that are inherent in the building operations, (2) the adequacy of the fire safety features currently located in the PFP Complex, and (3) the degree of compliance of the facility with specific fire safety provisions in DOE orders, related engineering codes, and standards.

  10. Extraction and use of historical extreme climate databases for nuclear power plants safety assessment

    Science.gov (United States)

    Hamdi, Yasser; Bertin, Xavier; Bardet, Lise; Duluc, Claire-Marie; Rebour, Vincent

    2015-04-01

    Safety assessments of nuclear power plants (NPPs) related to natural hazards are a matter of major interest to the nuclear community in France and many European countries. Over the past fewer decades, France has experienced many of these events such as heat waves (2003 and 2006), heavy snowstorms (1958, 1990 and 1992), storms which have given rise to heavy rain and severe floods (1992, 1999, 2010), strong straight-line wind and extreme marine surges (1987, 1999 and 2010) much larger than the other local observations (outliers). These outliers had clearly illustrated the potential to underestimate the extreme surges calculated with the current statistical methods. The estimation of extreme surges then requires the use of a statistical analysis approach having a more solid theoretical framework and using more reliable databases for the assessment of hazards to design NPPs to low or extremely low probabilities of failure. These databases can be produced by collecting historical information (HI) about severe climatic events occurred over short and long timescales. As a matter of fact, natural hazards such as heat waves, droughts, floods, severe storms and snowstorms have affected France and many European countries since the dawn of time. These events would have been such horrific experiences that if they really occurred, there would be unmistakable traces of them. They must have left clues. These catastrophic events have been unforgettably engraved in people's minds and many of them have been traced in archives and history textbooks. The oldest events have certainly left clues and traces somewhere in the geological layers of the earth or elsewhere. The construction of the historical databases and developing probabilistic approaches capable of integrating them correctly is highly challenging for the scientific community (Translating these geological clues to historical data to build historical databases that can be used by the statistical models is a different

  11. Vogtle Electric Generating Plant ETE Analysis Review

    Energy Technology Data Exchange (ETDEWEB)

    Diediker, Nona H.; Jones, Joe A.

    2006-12-09

    Under contract with the Nuclear Regulatory Commission (NRC), staff from Pacific Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL)-Albuquerque reviewed the evacuation time estimate (ETE) analysis dated April 2006 prepared by IEM for the Vogtle Electric Generating Plant (VEGP). The ETE analysis was reviewed for consistency with federal regulations using the NRC guidelines in Review Standard (RS)-002, Supplement 2 and Appendix 4 to NUREG-0654, and NUREG/CR-4831. Additional sources of information referenced in the analysis and used in the review included NUREG/CR-6863 and NUREG/CR-6864. The PNNL report includes general comments, data needs or clarifications, and requests for additional information (RAI) resulting from review of the ETE analysis.

  12. Plant safety margin against frost damages has declined in Switzerland over the last four decades

    Science.gov (United States)

    Vitasse, Yann; Schneider, Léonard; Klein, Geoffrey; Rixen, Christian; Rebetez, Martine

    2017-04-01

    Winters and early springs have become warmer over the last decades which has in turn promoted earlier plant development in temperate regions. While temperatures will on average continue to increase in the coming decades due to the rise of greenhouse gases concentration in the atmosphere, there is no consensus about how the occurrence of late spring frosts will change. If the frequency and the severity of late spring frosts remain unchanged in the future or advance less than vegetation onset, vulnerable plant organs (young leaves, flowers or dehardened buds) may be more exposed to frost damage. Here we analyzed long-term series of temperature data during the period 1975-2016 at 50 locations in Switzerland. We used different thresholds of growing degree days (GDD) as a proxy for spring phenology of fruit trees based on long-term series of phenological observations. Finally, we tested whether the time lag between the date when the GDD is reached and the latest occurrence of frost has changed over the study period. Overall we found that the safety margin against potential frost damage to plants has slightly decreased during the study period, irrespective of elevation (from 203 to 2283 m). Our results suggest that the cost for preventing frost damages on fruit trees could increase in the coming decades and the introduction of new varieties of fruit trees adapted to warmer climate should be carefully considered as they generally exhibit earlier spring phenology.

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

    Science.gov (United States)

    Rao, Suman

    2007-04-11

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, the 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. The Establishment of Object Selection Criteria for Effect Analysis of Electromagnetic Pulse (EMP) in Operating Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Song Hae; Ryu, Hosun; Kim, Minyi; Lee, Euijong [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The electromagnetic pulse (EMP) can be used as a strategic weapon by inducing damaging voltage and currents that the electrical circuits are not designed to withstand. EMPs are lethal to electronic systems. All EMP events have three common components: a source, coupling path, and receptor. It can also travel across power grids, destroying electronics as it passes in less than a second. There have been no research studies on the effect analysis for EMP in domestic nuclear power plants and power grids. To ensure the safety of operating nuclear power plants in this environment, the emission of EMP is needed for the effect analysis and safety measures against EMPs. Actually, it is difficult and inefficient to conduct the effect analysis of EMP with all the equipment and systems in nuclear power plants (NPPs). Therefore, this paper presents the results of establishing the object selection criteria for the effect analysis of EMP in operating nuclear power plants through reviewing previous research in the US and the safety related design concepts in domestic NPPs. It is not necessary to ensure the continued operation of the plant in intense multiple EMP environments. The most probable effect of EMP on a modern nuclear power plant is an unscheduled shutdown. EMP may also cause an extended shutdown by the unnecessary activation of some safety related systems. In general, EMP can be considered a nuisance to nuclear plants, but it is not considered a serious threat to plant safety. The results of EMP effect analysis show less possibility of failure in the tested individual equipment. It was also confirmed that there is no possibility of simultaneous failure for devices in charge of the safety shutdown in the NPP.

  15. Safety of virus-resistant transgenic plants two decades after their introduction: lessons from realistic field risk assessment studies.

    Science.gov (United States)

    Fuchs, Marc; Gonsalves, Dennis

    2007-01-01

    Potential safety issues have been raised with the development and release of virus-resistant transgenic plants. This review focuses on safety assessment with a special emphasis on crops that have been commercialized or extensively tested in the field such as squash, papaya, plum, grape, and sugar beet. We discuss topics commonly perceived to be of concern to the environment and to human health--heteroencapsidation, recombination, synergism, gene flow, impact on nontarget organisms, and food safety in terms of allergenicity. The wealth of field observations and experimental data is critically evaluated to draw inferences on the most relevant issues. We also express inside views on the safety and benefits of virus-resistant transgenic plants, and recommend realistic risk assessment approaches to assist their timely deregulation and release.

  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. Techniques for Analysis of Plant Phenolic Compounds

    Directory of Open Access Journals (Sweden)

    Thomas H. Roberts

    2013-02-01

    Full Text Available Phenolic compounds are well-known phytochemicals found in all plants. They consist of simple phenols, benzoic and cinnamic acid, coumarins, tannins, lignins, lignans and flavonoids. Substantial developments in research focused on the extraction, identification and quantification of phenolic compounds as medicinal and/or dietary molecules have occurred over the last 25 years. Organic solvent extraction is the main method used to extract phenolics. Chemical procedures are used to detect the presence of total phenolics, while spectrophotometric and chromatographic techniques are utilized to identify and quantify individual phenolic compounds. This review addresses the application of different methodologies utilized in the analysis of phenolic compounds in plant-based products, including recent technical developments in the quantification of phenolics.

  18. 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.

  19. 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.

  20. 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.

  1. 2014 PGSFR Safety Analysis for Loss of Heat Sink

    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)

    2015-10-15

    KAERI has been developing a conceptual design of the PGSFR (Prototype Gen-4 Sodium-cooled Fast Reactor) with the thermal power of 392.1 MWt, which is the pool type SFR (Sodium-cooled Fast Reactor) with metal fuel. The PGSFR consists of the PHTS (Primary Heat Transport System), the IHTS (Intermediate Heat Transport System), and the DHRS (Decay Heat Removal System). A LOHS (Loss Of Heat Sink) accident has been investigated for a safety evaluation of the PGSFR using the MARS-LMR code. The safety analysis is evaluated by a CDF (Cumulative Damage Function). In case of the LOHS accident, the tentative safety criterion is the CDF of under 0.05. The LOHS 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. Furthermore, it has been elucidated that LOHS with LOOP is more conservative than LOHS without LOOP.

  2. Response Time Evaluation for the Plant Protection System Using a Combined Technique of Analysis and Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Jae; Han, Seung; Yun, Jae Hee; Baek, Seung Min [KECO EnC, Inc., Daejeon (Korea, Republic of)

    2015-10-15

    This paper proposes the response time evaluation methodology for the plant protection system (PPS) trip channel for the advance power reactor 1400 (APR1400) nuclear power plant. To demonstrate that the PPS tip channel is functioning within its allowable response time limit, the proposed methodology uses the combined technique of both the response time analysis and test. The main purpose of determining the trip setpoint for safety systems is to meet the requirement of an analytical limit assumed in performing safety analyses. In addition, the response time assumed during safety analyses shall also be satisfied by the safety-related instrumentation. The response time is another critical factor required to ensure that the safety-related instrumentation channels accept the crucial assumptions of safety analyses. The response time evaluation methodology proposed herein is applied to the low steam generator level (LSGL) reactor trip parameter for the APR1400. The response time analysis for the LSGL trip parameter demonstrated that the analyzed response time would not exceed the allocated response time. The results of the response time also showed that all of the measured response times would be less than the analyzed response time.

  3. Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H. K.

    1981-10-01

    Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0

  4. Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hollnagel, E. [Linkoeping Univ. (Sweden). Dept. of Computer and Information Science; Gauthereau, V. [Linkoeping Univ. (Sweden). Dept. of Industrial Engineering

    2001-06-01

    interviews were conducted with technical staff at most of the Swedish NPPs. It focused on which solutions the various NPPs had developed to cope with the problem, and which steps had been taken specifically to improve the efficiency of ORV. It was soon found that ORV could not be separated from the rest of the work done in a NPP during outages since many of the proposed solutions have a broad scope. An analysis of the nine Swedish ORV cases had found weaknesses in four main areas: administration processes, management, human performance, and control room layout. Relative to these, the Swedish NPPs have implemented several technical and organisational solutions. Among the former are an overall re-qualification scheme, blocked safety functions, computerised operational position control, and central indications in the control room. Most of the technical solutions have been part of the design of the newer plants, since to implement them in older plants requires essential changes both in the station and in the control room. The organisational solutions comprised operational readiness plans, systematic ways of working, new instructions, co-ordinated testing, and the use of redundant or independent controls. Special emphasis was put on how the NPPs planned their outages, how the plans were implemented, and how deviations were handled. Issues related to learning from experience were also investigated. It was found that although all the NPPs approached the ORV issues in a serious and efficient manner, the solutions could be different corresponding to the characteristics of the organisation. Finally a number of questions, which still need answers, were identified. One is how new procedures or new barriers are accepted and assimilated into the safety culture. A second concerns the demarcation of systems for which ORV is required, i.e., the boundary between safety and non-safety systems. A third is how complex technical solutions influence the operators' work. Finally, it is

  5. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  6. Operational safety of turbine-generators at Loviisa nuclear power plant; Turbiini-generaattoreiden kaeyttoeturvallisuus Loviisan ydinvoimalaitoksella

    Energy Technology Data Exchange (ETDEWEB)

    Virolainen, T.

    1997-06-01

    The goal of the study is to assess the operational safety of the turbine-generators at the Loviisa NPP. The lay-out, operation, control, monitoring and testing of turbine-generators have been studied. Taking these findings into consideration and by using operational data of Loviisa and other power plants, the most significant safety issues of the turbine-generator system have been identified. The frequencies for initiating events and possible consequences have been determined based on plant operational experience and related literature. (58 refs.).

  7. An analysis of the impacts of economic incentive programs on commercial nuclear power plant operations and maintenance costs

    Energy Technology Data Exchange (ETDEWEB)

    Kavanaugh, D.C.; Monroe, W.H. [Pacific Northwest Lab., Richland, WA (United States); Wood, R.S. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-02-01

    Operations and Maintenance (O and M) expenditures by nuclear power plant owner/operators possess a very logical and vital link in considerations relating to plant safety and reliability. Since the determinants of O and M outlays are considerable and varied, the potential linkages to plant safety, both directly and indirectly, can likewise be substantial. One significant issue before the US Nuclear Regulatory Commission is the impact, if any, on O and M spending from state programs that attempt to improve plant operating performance, and how and to what extent these programs may affect plant safety and pose public health risks. The purpose of this study is to examine the role and degree of impacts from state promulgated economic incentive programs (EIPs) on plant O and M spending. A multivariate regression framework is specified, and the model is estimated on industry data over a five-year period, 1986--1990. Explanatory variables for the O and M spending model include plant characteristics, regulatory effects, financial strength factors, replacement power costs, and the performance incentive programs. EIPs are found to have statistically significant effects on plant O and M outlays, albeit small in relation to other factors. Moreover, the results indicate that the relatively financially weaker firms are more sensitive in their O and M spending to the presence of such programs. Formulations for linking spending behavior and EIPs with plant safety performance remains for future analysis.

  8. Assessment of Occupational Health and Safety for a Gas Meter Manufacturing Plant

    Science.gov (United States)

    Korkmaz, Ece; Iskender, Gulen; Germirli Babuna, Fatos

    2016-10-01

    This study investigates the occupational health and safety for a gas meter manufacturing plant. The risk assessment and management study is applied to plastic injection and mounting departments of the factory through quantitative Fine Kinney method and the effect of adopting 5S workplace organization procedure on risk assessment is examined. The risk assessment reveals that there are 17 risks involved; 14 grouped in high risk class (immediate improvement as required action); 2 in significant (measures to be taken as required action) and one in possible risk class (monitoring as required action). Among 14 high risks, 4 can be reduced by 83 % to be grouped under possible class when 5S is applied. One significant risk is observed to be lowered by 78 % and considered as possible risk due to the application of 5S. As a result of either 67 or 50 % reductions in 7 high risks, these risks are converted to be members of significant risk group after 5S implications.

  9. Efficacy and Safety of Medicinal Plants or Related Natural Products for Fibromyalgia: A Systematic Review

    Directory of Open Access Journals (Sweden)

    Simone de Souza Nascimento

    2013-01-01

    Full Text Available To assess the effects of medicinal plants (MPs or related natural products (RNPs on fibromyalgia (FM patients, we evaluate the possible benefits and advantages of MP or RNP for the treatment of FM based on eight randomized placebo-controlled trials (RCTs involving 475 patients. The methodological quality of all studies included was determined according to JADAD and “Risk of Bias” with the criteria in the Cochrane Handbook for Systematic Reviews of Interventions 5.1.0. Evidence suggests significant benefits of MP or RNP in sleep disruption, pain, depression, joint stiffness, anxiety, physical function, and quality of life. Our results demonstrated that MP or RNP had significant effects on improving the symptoms of FM compared to conventional drug or placebo; longer tests are required to determine the duration of the treatment and characterize the long-term safety of using MP, thus suggesting effective alternative therapies in the treatment of pain with minimized side effects.

  10. 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…

  11. Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G., E-mail: wrcc@cdtn.br, E-mail: egr@cdtn.br, E-mail: tanius@cdtn.br, E-mail: dhbs@cdtn.br, E-mail: tanius@cdtn.br, E-mail: raphaelmecanica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Souto, Joao P.R.S.; Carvalho Junior, Ideir T., E-mail: joprocha@yahoo.com.br, E-mail: ideir_engenharia@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Metalurgica

    2013-07-01

    Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

  12. How to evaluate the risks of work equipment and installations for health and safety? Research and activities of the German Committee for Plant Safety and consequences for regulation.

    Science.gov (United States)

    Pieper, R

    2012-01-01

    Work equipment and installations with a high risk for health and safety of employees should be paid a special attention. The German Product Safety Act, which is aimed to manufacturers or distributors in order to protect consumers, maintains a conclusive catalogue of these so-called "installations in need of monitoring" fixing the work equipment and installations for which such special inspections can be demanded. This catalogue has remained unchanged for decades and has been transformed nearly unmodified into the Plant Safety Ordinance. Currently, there is a discussion about this catalogue in Germany. A major point of concern is the definition and the significance of "especially" dangerous work equipment and installations. Two recent research projects are dealing with the problem how to define "especially".

  13. Advanced Power Plant Development and Analysis Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    A.D. Rao; G.S. Samuelsen; F.L. Robson; B. Washom; S.G. Berenyi

    2006-06-30

    Under the sponsorship of the U.S. Department of Energy/National Energy Technology Laboratory, a multi-disciplinary team led by the Advanced Power and Energy Program of the University of California at Irvine is defining the system engineering issues associated with the integration of key components and subsystems into advanced power plant systems with goals of achieving high efficiency and minimized environmental impact while using fossil fuels. These power plant concepts include 'Zero Emission' power plants and the 'FutureGen' H2 co-production facilities. The study is broken down into three phases. Phase 1 of this study consisted of utilizing advanced technologies that are expected to be available in the 'Vision 21' time frame such as mega scale fuel cell based hybrids. Phase 2 includes current state-of-the-art technologies and those expected to be deployed in the nearer term such as advanced gas turbines and high temperature membranes for separating gas species and advanced gasifier concepts. Phase 3 includes identification of gas turbine based cycles and engine configurations suitable to coal-based gasification applications and the conceptualization of the balance of plant technology, heat integration, and the bottoming cycle for analysis in a future study. Also included in Phase 3 is the task of acquiring/providing turbo-machinery in order to gather turbo-charger performance data that may be used to verify simulation models as well as establishing system design constraints. The results of these various investigations will serve as a guide for the U. S. Department of Energy in identifying the research areas and technologies that warrant further support.

  14. Reliability and safety of the K Reactor cooling system: Part 2, Engineering analysis of hydraulic and mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Shoemaker, R.H.

    1960-04-04

    Subsequent to the recent formulation and adoption of safety criteria for reactor cooling systems, there appeared the need for an independent evaluation of the safety and reliability of the K-Reactor cooling system in terms of these criteria. The primary, secondary and last-ditch cooling systems of this reactor involve a strong inter-dependence between electrical and hydraulic components of the water plant. Because of the complexity of inter-relationships between these components, the analysis was divided into two parallel studies which were accomplished during the simmer of 1959. F. D. Robbins has presented his analysis of the electrical power and control system in HW-61887. This report deals with an engineering analysis of the hydraulic and mechanical aspects of the reliability and safety of the K-Reactor Cooling System. The system, as described in this report, is that which existed during the simmer of 1959, prior to modification under Project CG-775 (now Project CG-883).

  15. KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop

    1997-07-01

    Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs.

  16. KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop

    1997-07-01

    Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs.

  17. Probabilistic safety goals for nuclear power plants; Phases 2-4. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bengtsson, L.; Knochenhauer, M. (Scandpower AB (Sweden)); Holmberg, J.-E.; Rossi, J. (VTT Technical Research Centre of Finland (Finland))

    2011-05-15

    Safety goals are defined in different ways in different countries and also used differently. Many countries are presently developing them in connection to the transfer to risk-informed regulation of both operating nuclear power plants (NPP) and new designs. However, it is far from self-evident how probabilistic safety criteria should be defined and used. On one hand, experience indicates that safety goals are valuable tools for the interpretation of results from a probabilistic safety assessment (PSA), and they tend to enhance the realism of a risk assessment. On the other hand, strict use of probabilistic criteria is usually avoided. A major problem is the large number of different uncertainties in a PSA model, which makes it difficult to demonstrate the compliance with a probabilistic criterion. Further, it has been seen that PSA results can change a lot over time due to scope extensions, revised operating experience data, method development, changes in system requirements, or increases of level of detail, mostly leading to an increase of the frequency of the calculated risk. This can cause a problem of consistency in the judgments. This report presents the results from the second, third and fourth phases of the project (2007-2009), which have dealt with providing guidance related to the resolution of some specific problems, such as the problem of consistency in judgement, comparability of safety goals used in different industries, the relationship between criteria on different levels, and relations between criteria for level 2 and 3 PSA. In parallel, additional context information has been provided. This was achieved by extending the international overview by contributing to and benefiting from a survey on PSA safety criteria which was initiated in 2006 within the OECD/NEA Working Group Risk. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by

  18. Incorporating organizational factors into probabilistic safety assessment of nuclear power plants through canonical probabilistic models

    Energy Technology Data Exchange (ETDEWEB)

    Galan, S.F. [Dpto. de Inteligencia Artificial, E.T.S.I. Informatica (UNED), Juan del Rosal, 16, 28040 Madrid (Spain)]. E-mail: seve@dia.uned.es; Mosleh, A. [2100A Marie Mount Hall, Materials and Nuclear Engineering Department, University of Maryland, College Park, MD 20742 (United States)]. E-mail: mosleh@umd.edu; Izquierdo, J.M. [Area de Modelado y Simulacion, Consejo de Seguridad Nuclear, Justo Dorado, 11, 28040 Madrid (Spain)]. E-mail: jmir@csn.es

    2007-08-15

    The {omega}-factor approach is a method that explicitly incorporates organizational factors into Probabilistic safety assessment of nuclear power plants. Bayesian networks (BNs) are the underlying formalism used in this approach. They have a structural part formed by a graph whose nodes represent organizational variables, and a parametric part that consists of conditional probabilities, each of them quantifying organizational influences between one variable and its parents in the graph. The aim of this paper is twofold. First, we discuss some important limitations of current procedures in the {omega}-factor approach for either assessing conditional probabilities from experts or estimating them from data. We illustrate the discussion with an example that uses data from Licensee Events Reports of nuclear power plants for the estimation task. Second, we introduce significant improvements in the way BNs for the {omega}-factor approach can be constructed, so that parameter acquisition becomes easier and more intuitive. The improvements are based on the use of noisy-OR gates as model of multicausal interaction between each BN node and its parents.

  19. Hazardous organic compounds in biogas plant end products--soil burden and risk to food safety.

    Science.gov (United States)

    Suominen, K; Verta, M; Marttinen, S

    2014-09-01

    The end products (digestate, solid fraction of the digestate, liquid fraction of the digestate) of ten biogas production lines in Finland were analyzed for ten hazardous organic compounds or compound groups: polychlorinated dibenzo-p-dioxins and furans (PCDD/Fs), polychlorinated biphenyls (PCB(7)), polyaromatic hydrocarbons (PAH(16)), bis-(2-ethylhexyl) phthalate (DEHP), perfluorinated alkyl compounds (PFCs), linear alkylbenzene sulfonates (LASs), nonylphenols and nonylphenol ethoxylates (NP+NPEOs), polybrominated diphenyl ethers (PBDEs), hexabromocyclododecane (HBCD) and tetrabromobisphenol A (TBBPA). Biogas plant feedstocks were divided into six groups: municipal sewage sludge, municipal biowaste, fat, food industry by-products, animal manure and others (consisting of milling by-products (husk) and raw former foodstuffs of animal origin from the retail trade). There was no clear connection between the origin of the feedstocks of a plant and the concentrations of hazardous organic compounds in the digestate. For PCDD/Fs and for DEHP, the median soil burden of the compound after a single addition of digestate was similar to the annual atmospheric deposition of the compound or compound group in Finland or other Nordic countries. For PFCs, the median soil burden was somewhat lower than the atmospheric deposition in Finland or Sweden. For NP+NPEOs, the soil burden was somewhat higher than the atmospheric deposition in Denmark. The median soil burden of PBDEs was 400 to 1000 times higher than the PBDE air deposition in Finland or in Sweden. With PBDEs, PFCs and HBCD, the impact of the use of end products should be a focus of further research. Highly persistent compounds, such as PBDE- and PFC-compounds may accumulate in agricultural soil after repeated use of organic fertilizers containing these compounds. For other compounds included in this study, agricultural use of biogas plant end products is unlikely to cause risk to food safety in Finland.

  20. Deterministic and risk-informed approaches for safety analysis of advanced reactors: Part II, Risk-informed approaches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Ahn, Sang Kyu; Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-05-15

    Technical insights and findings from a critical review of deterministic approaches typically applied to ensure design safety of nuclear power plants were presented in the companion paper of Part I included in this issue. In this paper we discuss the risk-informed approaches that have been proposed to make a safety case for advanced reactors including Generation-IV reactors such as Modular High-Temperature Gas-cooled Reactor (MHTGR), Pebble Bed Modular Reactor (PBMR), or Sodium-cooled Fast Reactor (SFR). Also considered herein are a risk-informed safety analysis approach suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently developed by the US Nuclear Regulatory Commission as a high-level regulatory infrastructure for safety evaluation of any type of reactor design. The insights from a comparative review of various deterministic and risk-informed approaches could be usefully used in developing a new licensing architecture for enhanced safety of evolutionary or advanced plants.

  1. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 2-Domino Hazard Index and case study.

    Science.gov (United States)

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

    2008-12-15

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design.

  2. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  3. 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.

  4. Analysis of safety from a human clinical trial with pterostilbene.

    Science.gov (United States)

    Riche, Daniel M; McEwen, Corey L; Riche, Krista D; Sherman, Justin J; Wofford, Marion R; Deschamp, David; Griswold, Michael

    2013-01-01

    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.

  5. 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.

  6. Accidental safety analysis methodology development in decommission of the nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, G. H.; Hwang, J. H.; Jae, M. S.; Seong, J. H.; Shin, S. H.; Cheong, S. J.; Pae, J. H.; Ang, G. R.; Lee, J. U. [Seoul National Univ., Seoul (Korea, Republic of)

    2002-03-15

    Decontamination and Decommissioning (D and D) of a nuclear reactor cost about 20% of construction expense and production of nuclear wastes during decommissioning makes environmental issues. Decommissioning of a nuclear reactor in Korea is in a just beginning stage, lacking clear standards and regulations for decommissioning. This work accident safety analysis in decommissioning of the nuclear facility can be a solid ground for the standards and regulations. For source term analysis for Kori-1 reactor vessel, MCNP/ORIGEN calculation methodology was applied. The activity of each important nuclide in the vessel was estimated at a time after 2008, the year Kori-1 plant is supposed to be decommissioned. And a methodology for risk analysis assessment in decommissioning was developed.

  7. Definition and means of maintaining the emergency notification and evacuation system portion of the plutonium finishing plant safety envelope

    Energy Technology Data Exchange (ETDEWEB)

    WHITE, W.F.

    1999-05-20

    The Emergency Evacuation and Notification System provides information to the Plutonium Finishing Plant (PFP) Building Emergency Director to assist in determining appropriate emergency response, notifies personnel of the required response, and assists in their response. The report identifies the equipment in the Safety Envelope (SE) for this System and the Administrative, Maintenance, and Surveillance Procedures used to maintain the SE Equipment.

  8. Pooling, meta-analysis, and the evaluation of drug safety

    Directory of Open Access Journals (Sweden)

    Leizorovicz Alain

    2002-03-01

    Full Text Available Abstract Background The "integrated safety report" of the drug registration files submitted to health authorities usually summarizes the rates of adverse events observed for a new drug, placebo or active control drugs by pooling the safety data across the trials. Pooling consists of adding the numbers of events observed in a given treatment group across the trials and dividing the results by the total number of patients included in this group. Because it considers treatment groups rather than studies, pooling ignores validity of the comparisons and is subject to a particular kind of bias, termed "Simpson's paradox." In contrast, meta-analysis and other stratified analyses are less susceptible to bias. Methods We use a hypothetical, but not atypical, application to demonstrate that the results of a meta-analysis can differ greatly from those obtained by pooling the same data. In our hypothetical model, a new drug is compared to 1 a placebo in 4 relatively small trials in patients at high risk for a certain adverse event and 2 an active reference drug in 2 larger trials of patients at low risk for this event. Results Using meta-analysis, the relative risk of experiencing the adverse event with the new drug was 1.78 (95% confidence interval [1.02; 3.12] compared to placebo and 2.20 [0.76; 6.32] compared to active control. By pooling the data, the results were, respectively, 1.00 [0.59; 1.70] and 5.20 [2.07; 13.08]. Conclusions Because these findings could mislead health authorities and doctors, regulatory agencies should require meta-analyses or stratified analyses of safety data in drug registration files.

  9. Valve inlet fluid conditions for pressurizer safety and relief valves in combustion engineering-designed plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bahr, J.; Chari, D.; Puchir, M.; Weismantel, S.

    1982-12-01

    The purpose of this study is to assemble documented information for C-E designed plants concerning pressurizer safety and power operated relief valve (PROV) inlet fluid conditions during actuation as calculated by conventional licensing analyses. This information is to be used to assist in the justification of the valve inlet fluid conditions selected for the testing of safety valves and PORVs in the EPRI/PWR Safety/Relief Valve Test Program. Available FSAR/Reload analyses and certain low temperature overpressurization analyses were reviewed to identify the pressurization transients which would actuate the valves, and the corresponding valve inlet fluid conditions. In addition, consideration was given to the Extended High Pressure Liquid Injection event. A general description of each pressurization transient is provided. The specific fluid conditions identified and tabulated for each C-E designed plant for each transient are peak pressurizer pressure, pressure ramp rate at actuation, temperature and fluid state.

  10. Development of Real Time Operating System for Safety Grade PLC (POSAFE-Q) for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Son, Han Seong [ENESYS, Taejon (Korea, Republic of); Hwang, Sung Jae [POSCON, Seoul (Korea, Republic of); Lee, Young Joon; Kim, Chang Hwoi; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    POSAFE-Q is a newly developed programmable logic controller (PLC) in order to apply to digital safety system of nuclear power plants (NPP) according to Nuclear Power Plant safety requirements. POSAFE-Q hardware and software development process, including design, review, verification and validation, and configuration control and quality assurance, satisfies the requirements imposed by 10CFR50, Appendix B. This article introduces a real time operating system pCOS, which is the core of POSAFE-Q. Section 2 describes the structure of pCOS. Section 3 describes a few important features of pCOS, which are necessary to the application for the digital safety system of NPP.0.

  11. Quantum analysis of a plant metacommunity

    Directory of Open Access Journals (Sweden)

    Kürşad Özkan

    2016-01-01

    Full Text Available A wide variety of statistical methods have been frequently used in the area of ecology. Especially probability based analytic approaches have been become more popular in the ecological studies in recent years. On the other hand, the core of quantum is based on calculation of energy footprint by probability approach. That is why it has been expected that quantum analysis will have found a significant place in the ecological studies in the future. As a result, the book of quantum ecology written by László Orlóci Frsc made a profound contribution to enter within this process. This study aims at calculations of the potential energy footprints of specific processes including phylogeny, environmental mediation and emergent effects by using quantum analysis for a plant metacommunity. In this study, the fundamental formulate was based on Max Planck’s energy based entropy modified by László Orlóci Frsc. A plant metacommunity example includes 3 complexes taken from various slope positions of Yazılı Canyon Nature Park located in the Lake districts was subjected to quantum analysis. Energy structure parameters were calculated for each complex. Variations among the complexes’ H(cx energy footprints were found insignificant. Calculations at the level of catena was shown that potential energy footprint of phylogeny overwhelms the effect of slope position in the nH terms. However this is turned in favor of the slope position gradient when the H footprints are compared. In the study, stability values of the complexes were also calculated and, the highest instability value was found for the complex taken from valley bottom of the canyon.

  12. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

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

  13. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

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

  14. Ares I-X Malfunction Turn Range Safety Analysis

    Science.gov (United States)

    Beaty, J. R.

    2011-01-01

    Ares I-X was the designation given to the flight test version of the Ares I rocket which was developed by NASA (also known as the Crew Launch Vehicle (CLV) component of the Constellation Program). The Ares I-X flight test vehicle achieved a successful flight test on October 28, 2009, from Pad LC-39B at Kennedy Space Center, Florida (KSC). As part of the flight plan approval for the test vehicle, a range safety malfunction turn analysis was performed to support the risk assessment and vehicle destruct criteria development processes. Several vehicle failure scenarios were identified which could have caused the vehicle trajectory to deviate from its normal flight path. The effects of these failures were evaluated with an Ares I-X 6 degrees-of-freedom (6-DOF) digital simulation, using the Program to Optimize Simulated Trajectories Version II (POST2) simulation tool. The Ares I-X simulation analysis provided output files containing vehicle trajectory state information. These were used by other risk assessment and vehicle debris trajectory simulation tools to determine the risk to personnel and facilities in the vicinity of the launch area at KSC, and to develop the vehicle destruct criteria used by the flight test range safety officer in the event of a flight test anomaly of the vehicle. The simulation analysis approach used for this study is described, including descriptions of the failure modes which were considered and the underlying assumptions and ground rules of the study.

  15. 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.

  16. Fast Flux Test Facility final safety analysis report. Amendment 73

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1993-08-01

    This report provides Final Safety Analysis Report (FSAR) Amendment 73 for incorporation into the Fast Flux Test Facility (FFTR) FSAR set. This page change incorporates Engineering Change Notices (ECNs) issued subsequent to Amendment 72 and approved for incorparoration before May 6, 1993. These changes include: Chapter 3, design criteria structures, equipment, and systems; chapter 5B, reactor coolant system; chapter 7, instrumentation and control systems; chapter 9, auxiliary systems; chapter 11, reactor refueling system; chapter 12, radiation protection and waste management; chapter 13, conduct of operations; chapter 17, technical specifications; chapter 20, FFTF criticality specifications; appendix C, local fuel failure events; and appendix Fl, operation at 680{degrees}F inlet temperature.

  17. Fast Flux Test Facility final safety analysis report. Amendment 73

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1993-08-01

    This report provides Final Safety Analysis Report (FSAR) Amendment 73 for incorporation into the Fast Flux Test Facility (FFTR) FSAR set. This page change incorporates Engineering Change Notices (ECNs) issued subsequent to Amendment 72 and approved for incorparoration before May 6, 1993. These changes include: Chapter 3, design criteria structures, equipment, and systems; chapter 5B, reactor coolant system; chapter 7, instrumentation and control systems; chapter 9, auxiliary systems; chapter 11, reactor refueling system; chapter 12, radiation protection and waste management; chapter 13, conduct of operations; chapter 17, technical specifications; chapter 20, FFTF criticality specifications; appendix C, local fuel failure events; and appendix Fl, operation at 680{degrees}F inlet temperature.

  18. Reliability analysis of wastewater treatment plants.

    Science.gov (United States)

    Oliveira, Sílvia C; Von Sperling, Marcos

    2008-02-01

    This article presents a reliability analysis of 166 full-scale wastewater treatment plants operating in Brazil. Six different processes have been investigated, comprising septic tank+anaerobic filter, facultative pond, anaerobic pond+facultative pond, activated sludge, upflow anaerobic sludge blanket (UASB) reactors alone and UASB reactors followed by post-treatment. A methodology developed by Niku et al. [1979. Performance of activated sludge process and reliability-based design. J. Water Pollut. Control Assoc., 51(12), 2841-2857] is used for determining the coefficients of reliability (COR), in terms of the compliance of effluent biochemical oxygen demand (BOD), chemical oxygen demand (COD), total suspended solids (TSS), total nitrogen (TN), total phosphorus (TP) and fecal or thermotolerant coliforms (FC) with discharge standards. The design concentrations necessary to meet the prevailing discharge standards and the expected compliance percentages have been calculated from the COR obtained. The results showed that few plants, under the observed operating conditions, would be able to present reliable performances considering the compliance with the analyzed standards. The article also discusses the importance of understanding the lognormal behavior of the data in setting up discharge standards, in interpreting monitoring results and compliance with the legislation.

  19. Acquisition and statistical analysis of reliability data for I and C parts in plant protection system

    Energy Technology Data Exchange (ETDEWEB)

    Lim, T. J.; Byun, S. S.; Han, S. H.; Lee, H. J.; Lim, J. S.; Oh, S. J.; Park, K. Y.; Song, H. S. [Soongsil Univ., Seoul (Korea)

    2001-04-01

    This project has been performed in order to construct I and C part reliability databases for detailed analysis of plant protection system and to develop a methodology for analysing trip set point drifts. Reliability database for the I and C parts of plant protection system is required to perform the detailed analysis. First, we have developed an electronic part reliability prediction code based on MIL-HDBK-217F. Then we have collected generic reliability data for the I and C parts in plant protection system. Statistical analysis procedure has been developed to process the data. Then the generic reliability database has been constructed. We have also collected plant specific reliability data for the I and C parts in plant protection system for YGN 3,4 and UCN 3,4 units. Plant specific reliability database for I and C parts has been developed by the Bayesian procedure. We have also developed an statistical analysis procedure for set point drift, and performed analysis of drift effects for trip set point. The basis for the detailed analysis can be provided from the reliability database for the PPS I and C parts. The safety of the KSNP and succeeding NPPs can be proved by reducing the uncertainty of PSA. Economic and efficient operation of NPP can be possible by optimizing the test period to reduce utility's burden. 14 refs., 215 figs., 137 tabs. (Author)

  20. Development of reliability-based safety enhancement technology; development of organization concept model in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Hyun; Kim, Ju Youl; Kim, Yoon Ik; Yang, Hui Chang; Lee, Yong Sik; Kim, Se Hyung [Seoul National University, Seoul (Korea)

    2002-03-01

    The influences of organizational factors on safety of nuclear power plants are mentioned in the early 1970s and noticed after being focused on in the accident report of TMI in 1979. These needs let us implement this research and the purposes of this research are to assess the organizational influences and to develop the organizational conceptual model to establish the basis of identifying the organizational factors, using this model to contribute to enhance safety and economics in nuclear power plants. Eventually research on the organizational influences is expected to have two effects, which are to improve safety through identifying potential causes of accidents and to elevate economics as a new approach to more efficient operation of nuclear power plants. In this study, recent studies were surveyed on the organizational conceptual model, the identification of organizational factors, assessment of organizational influences and evaluation methods of organizational factors and organizational influences among the overseas and domestic researches. In addition specific characteristics of domestic nuclear power plants were tried to identify through plant visit and an evaluation method of organizational influences on component maintenance and human performance were developed and presented. 71 refs., 40 figs., 18 tabs. (Author)

  1. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.

  2. A Difference-in-Differences Analysis of Health, Safety, and Greening Vacant Urban Space

    National Research Council Canada - National Science Library

    Branas, Charles C; Cheney, Rose A; MacDonald, John M; Tam, Vicky W; Jackson, Tara D; Ten Have, Thomas R

    2011-01-01

    Greening of vacant urban land may affect health and safety. The authors conducted a decade-long difference-in-differences analysis of the impact of a vacant lot greening program in Philadelphia, Pennsylvania, on health and safety outcomes...

  3. Nonlinear analysis of NPP safety against the aircraft attack

    Energy Technology Data Exchange (ETDEWEB)

    Králik, Juraj, E-mail: juraj.kralik@stuba.sk [Faculty of Civil Engineering, STU in Bratislava, Radlinského 11, 813 68 Bratislava (Slovakia); Králik, Juraj, E-mail: kralik@fa.stuba.sk [Faculty of Architecture, STU in Bratislava, Námestie Slobody 19, 812 45 Bratislava (Slovakia)

    2016-06-08

    The paper presents the nonlinear probabilistic analysis of the reinforced concrete buildings of nuclear power plant under the aircraft attack. The dynamic load is defined in time on base of the airplane impact simulations considering the real stiffness, masses, direction and velocity of the flight. The dynamic response is calculated in the system ANSYS using the transient nonlinear analysis solution method. The damage of the concrete wall is evaluated in accordance with the standard NDRC considering the spalling, scabbing and perforation effects. The simple and detailed calculations of the wall damage are compared.

  4. Nonlinear analysis of NPP safety against the aircraft attack

    Science.gov (United States)

    Králik, Juraj; Králik, Juraj

    2016-06-01

    The paper presents the nonlinear probabilistic analysis of the reinforced concrete buildings of nuclear power plant under the aircraft attack. The dynamic load is defined in time on base of the airplane impact simulations considering the real stiffness, masses, direction and velocity of the flight. The dynamic response is calculated in the system ANSYS using the transient nonlinear analysis solution method. The damage of the concrete wall is evaluated in accordance with the standard NDRC considering the spalling, scabbing and perforation effects. The simple and detailed calculations of the wall damage are compared.

  5. Probabilistic safety evaluation: Development of procedures with applications on components used in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Dillstroem, P. [Det Norske Veritas AB, Stockholm (Sweden)

    2000-12-01

    A probabilistic procedure has been developed by SAQ Kontroll AB to calculate two different failure probabilities, P{sub F}: Probability of failure, defect size given by NDT/NDE. Probability of failure, defect not detected by NDT/NDE. Based on the procedure, SAQ Kontroll AB has developed a computer program PROPSE (PRObabilistic Program for Safety Evaluation). Within PROPSE, the following features are implemented: Two different algorithms to calculate the probability of failure are included: Simple Monte Carlo Simulation (MCS), with an error estimate on P{sub F}. First-Order Reliability Method (FORM), with sensitivity factors using the most probable point of failure in a standard normal space. Using these factors, it is possible to rank the parameters within an analysis. Estimation of partial safety factors, given an input target failure probability and characteristic values for fracture toughness, yield strength, tensile strength and defect depth. Extensive validation has been carried out, using the probabilistic computer program STAR6 from Nuclear Electric and the deterministic program SACC from SAQ Kontroll AB. The validation showed that the results from PROPSE were correct, and that the algorithms used in STAR6 were not intended to work for a general problem, when the standard deviation is either 'small' or 'large'. Distributions, to be used in a probabilistic analysis, are discussed. Examples on data to be used are also given.

  6. Seismic analysis of nuclear power plant structures

    Science.gov (United States)

    Go, J. C.

    1973-01-01

    Primary structures for nuclear power plants are designed to resist expected earthquakes of the site. Two intensities are referred to as Operating Basis Earthquake and Design Basis Earthquake. These structures are required to accommodate these seismic loadings without loss of their functional integrity. Thus, no plastic yield is allowed. The application of NASTRAN in analyzing some of these seismic induced structural dynamic problems is described. NASTRAN, with some modifications, can be used to analyze most structures that are subjected to seismic loads. A brief review of the formulation of seismic-induced structural dynamics is also presented. Two typical structural problems were selected to illustrate the application of the various methods of seismic structural analysis by the NASTRAN system.

  7. Thermodynamic Analysis of Combined Cycle Power Plant

    Directory of Open Access Journals (Sweden)

    A.K.Tiwari,

    2010-04-01

    Full Text Available Air Bottoming Cycle (ABC can replace the heat recovery steam generator and the steam turbine of the conventionalcombined cycle plant. The exhaust energy of the topping gas turbine of existing combine cycle is sent to gas-air heat exchange, which heats the air in the secondary gas turbine cycle. In 1980’s the ABC was proposed as an alternative for the conventional steam bottoming cycle. In spite of the cost of reducing hardware installations it could achieve a thermal efficiency of 80%. The complete thermodynamic analysis of the system has been performed by using specially designed programme, enabling the variation of main independent variables. The result shows the gain in net work output as well as efficiency of combined cycle is 35% to 68%.

  8. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

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

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

    2007-04-11

    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.

  10. Improved analysis of bias in Monte Carlo criticality safety

    Science.gov (United States)

    Haley, Thomas C.

    2000-08-01

    Criticality safety, the prevention of nuclear chain reactions, depends on Monte Carlo computer codes for most commercial applications. One major shortcoming of these codes is the limited accuracy of the atomic and nuclear data files they depend on. In order to apply a code and its data files to a given criticality safety problem, the code must first be benchmarked against similar problems for which the answer is known. The difference between a code prediction and the known solution is termed the "bias" of the code. Traditional calculations of the bias for application to commercial criticality problems are generally full of assumptions and lead to large uncertainties which must be conservatively factored into the bias as statistical tolerances. Recent trends in storing commercial nuclear fuel---narrowed regulatory margins of safety, degradation of neutron absorbers, the desire to use higher enrichment fuel, etc.---push the envelope of criticality safety. They make it desirable to minimize uncertainty in the bias to accommodate these changes, and they make it vital to understand what assumptions are safe to make under what conditions. A set of improved procedures is proposed for (1) developing multivariate regression bias models, and (2) applying multivariate regression bias models. These improved procedures lead to more accurate estimates of the bias and much smaller uncertainties about this estimate, while also generally providing more conservative results. The drawback is that the procedures are not trivial and are highly labor intensive to implement. The payback in savings in margin to criticality and conservatism for calculations near regulatory and safety limits may be worth this cost. To develop these procedures, a bias model using the statistical technique of weighted least squares multivariate regression is developed in detail. Problems that can occur from a weak statistical analysis are highlighted, and a solid statistical method for developing the bias

  11. Integrated safety assessment of Indian nuclear power plants for extreme events: Reducing impact on public mind

    Indian Academy of Sciences (India)

    Anil Kakodkar; Ram Kumar Singh

    2013-10-01

    Nuclear energy professionals need to understand and address the catastrophe syndrome that of late seems to be increasingly at work in public mind in the context of nuclear energy. Classically the nuclear power reactor design and system evolution has been based on the logic of minimization of risk to an acceptable level and its quantification based on a deterministic approach and backed up by a further assessment based on the probabilistic methodology. However, in spite of minimization of risk, the reasons for anxiety and trauma in public mind that still prevails in the context of severe accidents needs to be understood and addressed. Margins between maximum credible accidents factored in the design and the ultimate load withstanding capacities of relevant systems need to be enhanced and guaranteed with a view to minimize release of radioactivity and avoid serious impact in public domain. A more realistic basis for management of an accident in public domain also needs to be quantified for this purpose. Assurance to public on limiting the consequences to a level that does not lead to a trauma is something that we need to be able to credibly demonstrate and confirm. The findings from Chernobyl reports point to significant psychological effects and related health disorders due to large scale emergency relocation of people that could have been possibly reduced by an order of magnitude without significant additional safety detriment. A combination of probabilistic and deterministic approaches should be evolved further to minimize consequences in public domain through enhancing safety margins and adding greater precision to quantitatively predicting accident progression and its management. The paper presents the case studies of the extreme external event such as tsunami and its impact on the coastal nuclear plants in India, the containment integrity assessment under the extreme internal event of over-pressurization and aircraft impact along with hydrogen deflagration

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

    Science.gov (United States)

    Withrow, Colleen A.; Reveley, Mary S.

    2014-01-01

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

  13. A De Novo-Assembly Based Data Analysis Pipeline for Plant Obligate Parasite Metatranscriptomic Studies

    OpenAIRE

    Li Guo; Kelly S Allen; Greg A Deiulio; Yong Zhang; Madeiras, Angela M.; Robert L Wick; Li-Jun Ma

    2016-01-01

    Current and emerging plant diseases caused by obligate parasitic microbes such as rusts, downy mildews, and powdery mildews threaten worldwide crop production and food safety. These obligate parasites are typically unculturable in the laboratory, posing technical challenges to characterize them at the genetic and genomic level. Here we have developed a data analysis pipeline integrating several bioinformatic software programs. This pipeline facilitates rapid gene discovery and expression anal...

  14. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  15. Plant databases and data analysis tools

    Science.gov (United States)

    It is anticipated that the coming years will see the generation of large datasets including diagnostic markers in several plant species with emphasis on crop plants. To use these datasets effectively in any plant breeding program, it is essential to have the information available via public database...

  16. Documented Safety Analysis for the B695 Segment

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-09-11

    This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, or {sup 3}H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building

  17. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-20

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D&D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities.

  18. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  19. A 3-year hygiene and safety monitoring of a meat processing plant which uses raw materials of global origin.

    Science.gov (United States)

    Manios, Stavros G; Grivokostopoulos, Nikolaos C; Bikouli, Vasiliki C; Doultsos, Dimitrios A; Zilelidou, Evangelia A; Gialitaki, Maria A; Skandamis, Panagiotis N

    2015-09-16

    A systematic approach in monitoring the hygiene of a meat processing plant using classical microbiological analyses combined with molecular characterization tools may assist in the safety of the final products. This study aimed: (i) to evaluate the total hygiene level and, (ii) to monitor and characterize the occurrence and spread of Salmonella spp. and Listeria monocytogenes in the environment and the final products of a meat industry that processes meat of global origin. In total, 2541 samples from the processing environment, the raw materials, and the final products were collected from a Greek meat industry in the period 2011-2013. All samples were subjected to enumeration of total viable counts (TVC), Escherichia coli (EC) and total coliforms (TCC) and the detection of Salmonella spp., while 709 of these samples were also analyzed for the presence L. monocytogenes. Pathogen isolates were serotyped and further characterized for their antibiotic resistance and subtyped by PFGE. Raw materials were identified as the primary source of contamination, while improper handling might have also favored the proliferation of the initial microbial load. The occurrence of Salmonella spp. and L. monocytogenes reached 5.5% and 26.9%, respectively. Various (apparent) cross-contamination or persistence trends were deduced based on PFGE analysis results. Salmonella isolates showed wide variation in their innate antibiotic resistance, contrary to L. monocytogenes ones, which were found susceptible to all antibiotics except for cefotaxime. The results emphasize the biodiversity of foodborne pathogens in a meat industry and may be used by meat processors to understand the spread of pathogens in the processing environment, as well as to assist the Food Business Operator (FBO) in establishing effective criteria for selection of raw materials and in improving meat safety and quality. This approach can limit the increase of microbial contamination during the processing steps observed in

  20. Survey of systems safety analysis methods and their application to nuclear waste management systems

    Energy Technology Data Exchange (ETDEWEB)

    Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

  1. Canister storage building (CSB) safety analysis report phase 3: Safety analysis documentation supporting CSB construction

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1997-04-28

    The Canister Storage Building (CSB) will be constructed in the 200 East Area of the U.S. Department of Energy (DOE) Hanford Site. The CSB will be used to stage and store spent nuclear fuel (SNF) removed from the Hanford Site K Basins. The objective of this chapter is to describe the characteristics of the site on which the CSB will be located. This description will support the hazard analysis and accident analyses in Chapter 3.0. The purpose of this report is to provide an evaluation of the CSB design criteria, the design's compliance with the applicable criteria, and the basis for authorization to proceed with construction of the CSB.

  2. Safety related analysis of the application and operation of electrical components in German nuclear power plants, safeguarding and protection against safety relevant impacts from the grid and other external sources; Sicherheitstechnische Analyse zum Einsatz und Betrieb elektrotechnischer Einrichtungen in deutschen Kernkraftwerken, Ueberwachung und Schutz gegen sicherheitstechnisch bedeutsame Einwirkungen aus dem Verbundnetz sowie anderen aeusseren Quellen

    Energy Technology Data Exchange (ETDEWEB)

    Arians, Robert; Arnold, Simone; Blum, Stefanie; Buchholz, Marcel; Lochthofen, Andre; Quester, Claudia; Sommer, Dagmar

    2015-10-15

    In this report, results and data from examinations concerning software-based electrical components and transmitters are evaluated. As failure modes of software-based com-ponents and failure causes differ fundamentally from non-software-based components, an evaluation of the operating experience of such components was carried out. This evaluation should show whether or not existing approaches for non-software-based components can be directly transferred to software-based components, or if a different approach has to be developed. To include failures in non-safety systems, events not fulfilling the incident reporting criteria of German authorities were also included in this evaluation. The data provided by licensees of six German NPPs (different Boiling Wa-ter Reactors and Pressurized Water Reactors) was recorded for at least 8 years. The software-based components used in the NPPs are identified and their operating experience is analyzed in order to identify relevant failure modes and to establish a II knowledge base for future failure rating. In addition, the state of the art and science concerning the specific components was described.

  3. Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation.

  4. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

    1999-07-01

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

  5. Environment, safety and Health Progress Assessment of the Rocky Flats Plant

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    This report documents the result of the US Department of Energy`s (DOE) Environment, Safety and Health (ES&H) Progress Assessment of the DOE Rocky Flats Plant (RFP) in Golden, Colorado. The assessment, which was conducted during the period of May 17 through May 28, 1993, included a selective review of the ES&H management systems and programs of the responsible DOE Headquarters Program Offices (Defense Programs (DP) and Environmental Restoration and Waste Management (EM)), the DOE Rocky Flats Office (RFO), and the site contractor, EG&G Rocky Flats, Inc. (EG&G). Despite the near constant state of flux under which RFP has been required to operate, the Progress Assessment Team has concluded that significant progress has been made in correcting the deficiencies identified in the 1989 Assessment and in responding responsibly to regulations, and DOE directives and guidance that have been issued since that time. The Team concluded that the improvements have been concentrated in the activities associated with plutonium facilities and in regulatory driven programs. Much remains to be done with respect to implementing on a sitewide basis those management systems that anchor an organization`s pursuit of continuous ES&H improvement. Furthermore the Team concluded that the pace of improvement has been constrained by a combination of factors that have limited the site`s ability to manage change in the pursuit of sitewide ES&H excellence.

  6. The risk analysis of dust electrostatic based on on-site survey of polypropylene plant

    Science.gov (United States)

    Wu, Xiumin; He, Mingjun; Yu, Haibo

    2013-03-01

    The dust electrostatic explosion accidents in polypropylene plant are mainly caused by the interaction of combustible gas, dust and static electricity. This paper analyses the key parts easy to produce dust and the risks of dust electrostatic by on-site survey of polypropylene plant, and proposes corresponding safety protection measures. The analysis results indicate that any careless mistakes and deviation in every step of process control may lead to electrostatic explosion in the silo. And if the equipment has some inherent defects and there are some careless mistakes in the process control, it will be easier to cause dust electrostatic explosion accidents.

  7. SAFETY-BASED CAPACITY ANALYSIS FOR CHINESE HIGHWAYS

    Directory of Open Access Journals (Sweden)

    Ping YI, Ph.D.

    2004-01-01

    Full Text Available Many years of research have led to the development of theories and methodologies in roadway capacity analysis in the developed countries. However, those resources coexist with roadway design and traffic control practices in the local country, and cannot be simply transferred to China for applications. For example, the Highway Capacity Manual in the United State describes roadway capacity under ideal conditions and estimates practical capacities under prevailing conditions in the field. This capacity and the conditions for change are expected to be different on Chinese roadways as the local roadway design (lane width, curves and grades, vehicle size, and traffic mix are different. This research looks into an approach to the capacity issue different from the Highway Capacity Manual. According to the car-following principle, this paper first describes the safety criteria that affect traffic operations. Several speed schemes are subsequently discussed as they are affected by the maximum speed achievable under the local conditions. The study has shown that the effect of geometric and traffic conditions can be effectually reflected in the maximum speed adopted by the drivers. For most Chinese highways without a posted speed limit, the choice of speed by the drivers from the safety prospective is believed to have incorporated considerations of the practical driving conditions. Based on this, a condition for capacity calculation is obtained by comparing the desired vs. safety-based distance headways. The formulations of the model are mathematically sound and physically meaningful, and preliminary testing of the model is encouraging. Future research includes field data acquisition for calibration and adjustment, and model testing on Chinese highways.

  8. Hazard Analysis and Safety Requirements for Small Drone Operations: To What Extent Do Popular Drones Embed Safety?

    Science.gov (United States)

    Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimihailidou, Maria Mikela

    2017-08-02

    Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this article presents: (1) a set of safety requirements generated from the application of the systems theoretic process analysis (STPA) technique on a generic small drone system; (2) a gap analysis between the set of safety requirements and the ones met by 19 popular drone models; (3) the extent of the differences between those models, their manufacturers, and the countries of origin; and (4) the association of drone prices with the extent they meet the requirements derived by STPA. The application of STPA resulted in 70 safety requirements distributed across the authority, manufacturer, end user, or drone automation levels. A gap analysis showed high dissimilarities regarding the extent to which the 19 drones meet the same safety requirements. Statistical results suggested a positive correlation between drone prices and the extent that the 19 drones studied herein met the safety requirements generated by STPA, and significant differences were identified among the manufacturers. This work complements the existing risk assessment frameworks for small drones, and contributes to the establishment of a commonly endorsed international risk analysis framework. Such a framework will support the development of a holistic and methodologically justified standardization scheme for small drone flights. © 2017 Society for Risk Analysis.

  9. A Study on the Adverse Effect of AOVs in AFWS Recirculation Paths on Plant Safety

    Energy Technology Data Exchange (ETDEWEB)

    Huong, Ho Thi Thanh [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chung, Dae-Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    Auxiliary feedwater system (AFWS) recirculation flow paths adopted air operated valves (AOVs) which could fail close on loss of instrument air (LOIA) event. So the AOVs and recirculation paths are closed on LOIA event, which could result in Auxiliary feedwater (AFW) pump(s) damage, which contributes greatly to core damage frequency (CDF).. On February 2002, the USNRC issued an inspection finding related to potential common cause failure of AOVs in AFWS recirculation flow paths on loss of instrument air system in Point Beach nuclear power plant (Pt. Beach). The AOVs have been removed from AFWS recirculation paths in the design of Korea standard nuclear power plant (KSNP). So, there is no possibility of above mentioned failure event in KSNP. It would be beneficial to evaluate the significance of adverse effect of AOVs in AFWS recirculation paths to realize the importance of maintaining AFWS recirculation paths always open. In this study, the AFWS modeling of Ulchin unit 3 and 4 was modified to model the AOVs in AFW recirculation flow paths to evaluate the change in CDF, which is caused by the adverse effect of AOV with operation mode of 'fails close' on LOIA event. It is concluded that the existence of AOV with 'fail close' design in AFWS MDP recirculation paths results in CDF increase of 131%, which is significant adverse effect on plant safety.. In this regard, the improved Westinghouse design and KSNP design had removed the AOVs from AFWS MDP recirculation paths. However, a couple of units with old Westinghouse design, Kori 1 and 2, still have AOVs in AFWS MDP recirculation paths and throttle back operation of AFWS is in effect. Although those AOVs adopt 'fail open' design to prevent above mentioned inadvertent closure, considering the big increase in CDF, there still exists considerable risk from the possibility of 'failure to open' during this throttle back operation. Therefore, it is strongly recommended that any

  10. Procedure for conducting probabilistic safety assessment: level 1 full power internal event analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Won Dae; Lee, Y. H.; Hwang, M. J. [and others

    2003-07-01

    This report provides guidance on conducting a Level I PSA for internal events in NPPs, which is based on the method and procedure that was used in the PSA for the design of Korea Standard Nuclear Plants (KSNPs). Level I PSA is to delineate the accident sequences leading to core damage and to estimate their frequencies. It has been directly used for assessing and modifying the system safety and reliability as a key and base part of PSA. Also, Level I PSA provides insights into design weakness and into ways of preventing core damage, which in most cases is the precursor to accidents leading to major accidents. So Level I PSA has been used as the essential technical bases for risk-informed application in NPPs. The report consists six major procedural steps for Level I PSA; familiarization of plant, initiating event analysis, event tree analysis, system fault tree analysis, reliability data analysis, and accident sequence quantification. The report is intended to assist technical persons performing Level I PSA for NPPs. A particular aim is to promote a standardized framework, terminology and form of documentation for PSAs. On the other hand, this report would be useful for the managers or regulatory persons related to risk-informed regulation, and also for conducting PSA for other industries.

  11. Nuclear power plant personnel errors in decision-making as an object of probabilistic risk assessment. Methodological extensions on the basis of a differentiated analysis of safety-relevant goals; Entscheidungsfehler des Betriebspersonals von Kernkraftwerken als Objekt probabilistischer Risikoanalysen; Methodische Erweiterungen auf der Basis einer differenzierten Betrachtungsweise sicherheitsgerichteter Ziele

    Energy Technology Data Exchange (ETDEWEB)

    Reer, B.

    1993-09-01

    Integration of human error (man-machine system analysis (MMSA)) is an essential part of probabilistic risk assessment (PRA). A method is presented for systematic, comprehensive PRA inclusions of decision-based errors due to conflicts or similarities. For error identification procedure, new question techniques are developed. These errors are identified by looking at retroactions caused by subordinate goals as components of overall safety relevant goal. New quantification methods for estimating situation-specific probabilities are developed. The factors conflict and similarity are operationalized in a way that allows their quantification based on informations usually available in PRA. Quantification procedure uses extrapolations and interpolations based on a poor set of data related to decision-based errors. Moreover, for passive errors in decision-making a completely new approach is presented where errors are quantified via a delay initiating the required action rather than via error probabilities. Practicability of this dynamic approach is demonstrated by probabilistic analysis of the actions required during the total loss of feedwater event at the Davis-Besse plant 1985. The extensions of the classical PRA method developed in this work are applied to a MMSA of the decay heat removal (DHR) of the HTR-500. Errors in decision-making - as potential roots of extraneous acts - are taken into account in a comprehensive and systematic manner. Five additional errors are identified. However, the probabilistic quantification results a nonsignificant increase of the DHR failure probability. (orig.) [Deutsch] Einbeziehung von Operateurfehlern (Mensch-Maschine-Systemanalyse (MMSA)) ist Bestandteil einer probabilistischen Risikoanalyse (PRA). Es wird eine Methode vorgestellt, mit der sich Entscheidungsfehler aufgrund der Faktoren Konflikt und Aehnlichkeit systematisch und umfassend in MMSA integrieren lassen. Zur Identifizierung der entsprechenden Situationen im Stoerfallablauf

  12. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  13. Safety - a Neglected Issue When Introducing Solid Biomass Fuel in Thermal Power Plants? Some Evidence of an Emerging Risk

    DEFF Research Database (Denmark)

    Hedlund, Frank Huess; Astad, John

    2013-01-01

    The paper examines recent evidence from Denmark and abroad with climate change projects that aim to reduce global carbon dioxide emissions by converting coal fired thermal power plants to solid biomass fuel. The paper argues that projects appear to be pursued narrow-mindedly with insufficient...... attention paid to safety and points to evidence of media-shifting-that the 'resolution' of a problem within the environmental domain creates a new problem in the workplace safety domain. The paper argues that biomass pellets qualify as an emerging risk for which proper control strategies have yet...

  14. Radiation safety issues in the water treatment plant - Indoor radon and gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Jantsikene, A.; Kiisk, M.; Suursoo, S.; Koch, R. [University of Tartu, Institute of Physics (Estonia); Lumiste, L. [Tallinn University of Technology, Department of Chemical Engineering (Estonia)

    2014-07-01

    the second stage filters, classifying them as radioactive material. {sup 222}Rn activity generated in the filter columns was compared to {sup 222}Rn activity entering the plant with raw water; the latter was estimated via liquid scintillation technique. The results indicate that {sup 222}Rn inflow is insignificant compared to the {sup 222}Rn generation in the columns. The estimated difference is in six orders of magnitude. Six alpha-track detectors were placed inside the treatment plant (in filter hall and working rooms) for measuring {sup 220}Rn and {sup 222}Rn indoor air concentrations. The obtained results are above 200 Bq/m{sup 3}. According to the Estonian Standard EVS 840:2009 'Design of radon-safe buildings', the annual radon concentration in living-, rest- and workrooms should remain below 200 Bq/m{sup 3}. The implemented investigation presented in this paper reveals the importance of monitoring radiation safety of the workers of this water treatment plant. Document available in abstract form only. (authors)

  15. Documented Safety Analysis for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  16. Documented Safety Analysis for the Waste Storage Facilities March 2010

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  17. Hazard screening application guide. Safety Analysis Report Update Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-06-01

    The basic purpose of hazard screening is to group precesses, facilities, and proposed modifications according to the magnitude of their hazards so as to determine the need for and extent of follow on safety analysis. A hazard is defined as a material, energy source, or operation that has the potential to cause injury or illness in human beings. The purpose of this document is to give guidance and provide standard methods for performing hazard screening. Hazard screening is applied to new and existing facilities and processes as well as to proposed modifications to existing facilities and processes. The hazard screening process evaluates an identified hazards in terms of the effects on people, both on-site and off-site. The process uses bounding analyses with no credit given for mitigation of an accident with the exception of certain containers meeting DOT specifications. The process is restricted to human safety issues only. Environmental effects are addressed by the environmental program. Interfaces with environmental organizations will be established in order to share information.

  18. Nuclear Safety Analysis for the Mars Exploration Rover 2003 Project

    Science.gov (United States)

    Firstenberg, Henry; Rutger, Lyle L.; Mukunda, Meera; Bartram, Bart W.

    2004-02-01

    The National Aeronautics and Space Administration's Mars Exploration Rover (MER) 2003 project is designed to place two mobile laboratories (Rovers) on Mars to remotely characterize a diversity of rocks and soils. Milestones accomplished so far include two successful launches of identical spacecraft (the MER-A and MER-B missions) from Cape Canaveral Air Force Station, Florida on June 10 and July 7, 2003. Each Rover uses eight Light Weight Radioisotope Heater Units (LWRHUs) fueled with plutonium-238 dioxide to provide local heating of Rover components. The LWRHUs are provided by the U.S. Department of Energy. In addition, small quantities of radioactive materials in sealed sources are used in scientific instrumentation on the Rover. Due to the radioactive nature of these materials and the potential for accidents, a formal Launch Approval Process requires the preparation of a Final Safety Analysis Report (FSAR) for submittal to and independent review by an Interagency Nuclear Safety Review Panel. This paper presents a summary of the FSAR in terms of potential accident scenarios, probabilities, source terms, radiological consequences, mission risks, and uncertainties in the reported results.

  19. Large break loss-of-coolant accident analysis for China Qinshan-2 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Ban, Chang Hwan; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Wang, Rongzhong; Yu, Hongxing [Nuclear Power Institute of China, Chengdu, SC (China)

    1994-12-01

    Large break LOCA analysis for China Qinshan-2 nuclear power plant has been performed using realistic evaluation model which has been being developed by KAERI. RELAP5/MOD3/KAERI code, which is a modified version of RELAP5/MOD3, is coupled with CONTEMPT4/MOD5 and is used as a best estimate code to predict the thermal hydraulic behavior of the system. PCT uncertainty which stems from code uncertainty, plant application uncertainty, scaling uncertainty and PCT bias are discussed. Among them, plant application uncertainty is described in detail. The licensing PCT is calculated by adding all the uncertainties to the best-estimate PCT. The result indicates the Qinshan-2 nuclear power plant has at least 37 deg C safety margin for large break LOCA. (Author) 10 refs., 47 figs., 14 tabs.

  20. Integrated Safety Management System Phase I Verification for the Plutonium Finishing Plant (PFP) [VOL 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    SETH, S.S.

    2000-01-10

    U.S. Department of Energy (DOE) Policy 450.4, Safety Management System Policy commits to institutionalizing an Integrated Safety Management System (ISMS) throughout the DOE complex as a means of accomplishing its missions safely. DOE Acquisition Regulation 970.5204-2 requires that contractors manage and perform work in accordance with a documented safety management system.