WorldWideScience

Sample records for safety calculations involving

  1. Validation of KENO V.a for criticality safety calculations involving WR-1 fast-neutron fuel arrangements

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.

    1991-07-15

    The KENO V.a criticality safety code, used with the SCALE 27-energy-group ENDF/B-IV-based cross-section library, has been validated for low-enriched uranium carbide (UC) WR-1 fast-neutron (FN) fuel arrangements. Because of a lack of relevant experimental data for UC fuel in the published literature, the validation is based primarily on calculational comparisons with critical experiments for fuel types with a range of enrichments and densities that cover those of the FN UC fuel. The ability of KENO V.a to handle the unique annular pin arrangement of the WR-1 FN fuel bundle was established using a comparison with the MCNP3B code used with a continuous-energy ENDF/B-V-based cross-section library. This report is part of the AECL--10146 report series documenting the validation of the KENO V.a criticality safety code.

  2. Calculation and definition of safety indicators

    International Nuclear Information System (INIS)

    Cristian, I.; Branzeu, N.; Vidican, D.; Vladescu, G.

    1997-01-01

    This paper presents, based on Cernavoda safety indicators proposal, the purpose definition and calculation formulas for each of the selected safety indicators. Five categories of safety indicators for Cernavoda Unit 1 were identified, namely: overall plant safety performance; initiating events; safety system availability, physical barrier integrity; indirect indicators. Definition, calculation and use of some safety indicators are shown in a tabular form. (authors)

  3. Safety Climate, Perceived Risk, and Involvement in Safety Management

    OpenAIRE

    Kouabenan , Dongo Rémi; Ngueutsa , Robert ,; Safiétou , Mbaye

    2015-01-01

    International audience; This article examines the relationship between safety climate, risk perception and involvement in safety management by first-line managers (FLM). Sixty-three FLMs from two French nuclear plants answered a questionnaire measuring perceived workplace safety climate, perceived risk, and involvement in safety management. We hypothesized that a positive perception of safety climate would promote substantial involvement in safety management, and that this effect would be str...

  4. Practicing industrial safety - issues involved

    International Nuclear Information System (INIS)

    Gunasekaran, P.

    2016-01-01

    Industrial safety is all about measures or techniques implemented to reduce the risk of injury, loss to persons, property or the environment in any industrial facility. The issue of industrial safety evolved concurrently with industrial development as a shift from compensation to prevention as well. Today, industrial safety is widely regarded as one of the most important factors that any business, large or small, must consider in its operations, as prevention of loss is also a part of profit. Factories Act of Central government and Rules made under it by the state deals with the provisions on industrial safety legislation. There are many other acts related to safety of personnel, property and environment. Occupational health and safety is also of primary concern. The aim is to regulate health and safety conditions for all employers. It includes safety standards and health standards. These acts encourage employers and employees to reduce workplace hazards and to implement new or improve existing safety and health standards; and develop innovative ways to achieve them. Maintain a reporting and record keeping system to monitor job-related injuries and illnesses; establish training programs to increase the number and competence of occupational safety and health personnel

  5. Criticality calculations for safety analysis

    International Nuclear Information System (INIS)

    Vellozo, S.O.

    1981-01-01

    Criticality studies in uranium nitrate and plutonium nitrate aqueous solutions were done. For uranium compound three basic computer codes are used: GAMTEC-II, DTF-IV, KENO-IV. Water was used as refletor and the results obtained with the different computer codes were analyzed and compared with the 'Handbuck zur Kriticalitat'. The cross sections and the cylindrical geometry were generated by Gamtec-II computer code. In the second compound the thickness of the recipient with plutonium nitrate are used with rectangular geometry and concret reflector. The effective multiplication constant was calculated with the Gamtec-II and Keno-IV library. The results show many differences. (E.G) [pt

  6. DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

  7. Benchmarking criticality safety calculations with subcritical experiments

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1984-06-01

    Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

  8. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  9. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  10. Temperature calculation in fire safety engineering

    CERN Document Server

    Wickström, Ulf

    2016-01-01

    This book provides a consistent scientific background to engineering calculation methods applicable to analyses of materials reaction-to-fire, as well as fire resistance of structures. Several new and unique formulas and diagrams which facilitate calculations are presented. It focuses on problems involving high temperature conditions and, in particular, defines boundary conditions in a suitable way for calculations. A large portion of the book is devoted to boundary conditions and measurements of thermal exposure by radiation and convection. The concepts and theories of adiabatic surface temperature and measurements of temperature with plate thermometers are thoroughly explained. Also presented is a renewed method for modeling compartment fires, with the resulting simple and accurate prediction tools for both pre- and post-flashover fires. The final chapters deal with temperature calculations in steel, concrete and timber structures exposed to standard time-temperature fire curves. Useful temperature calculat...

  11. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    Science.gov (United States)

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  12. Methodology for calculating guideline concentrations for safety shot sites

    International Nuclear Information System (INIS)

    1997-06-01

    Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination

  13. Methodology for calculating guideline concentrations for safety shot sites

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination.

  14. Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

    Energy Technology Data Exchange (ETDEWEB)

    Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-26

    The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

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

    International Nuclear Information System (INIS)

    Van Keuren, J.C.

    1997-01-01

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

  16. Kowledge-based dynamic network safety calculations. Wissensbasierte dynamische Netzsicherheitsberechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Kulicke, B [Inst. fuer Hochspannungstechnik und Starkstromanlagen, Berlin (Germany); Schlegel, S [Inst. fuer Hochspannungstechnik und Starkstromanlagen, Berlin (Germany)

    1993-06-28

    An important part of network operation management is the estimation and maintenance of the security of supply. So far the control personnel has only been supported by static network analyses and safety calculations. The authors describe an expert system, which is coupled to a real time simulation program on a transputer basis, for dynamic network safety calculations. They also introduce the system concept and the most important functions of the expert system. (orig.)

  17. NPP Krsko core calculations to improve operational safety

    International Nuclear Information System (INIS)

    Ivekovic, I.; Grgic, D.; Nemec, T.

    2007-01-01

    Calculation tools and methodology used to perform independent calculations of cumulative influence of different changes related to fuel and core operation of NPP Krsko were described. Some examples of steady state and transient results are used to illustrate potential improvements to understanding and reviewing plant safety. (author)

  18. Validation of calculational methods for nuclear criticality safety - approved 1975

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, N16.1-1975, states in 4.2.5: In the absence of directly applicable experimental measurements, the limits may be derived from calculations made by a method shown to be valid by comparison with experimental data, provided sufficient allowances are made for uncertainties in the data and in the calculations. There are many methods of calculation which vary widely in basis and form. Each has its place in the broad spectrum of problems encountered in the nuclear criticality safety field; however, the general procedure to be followed in establishing validity is common to all. The standard states the requirements for establishing the validity and area(s) of applicability of any calculational method used in assessing nuclear criticality safety

  19. Review of safety reports involving electronic flight bags

    Science.gov (United States)

    2009-04-27

    Electronic Flight Bags (EFBs) are a relatively new device used by pilots. Even so, 37 safety-related events involving EFBs were identified from the public online Aviation Safety Reporting System (ASRS) database as of June 2008. In addition, two accid...

  20. Calculational study for criticality safety data of fissionable actinides

    International Nuclear Information System (INIS)

    Nojiri, Ichiro; Fukasaku, Yasuhiro.

    1997-01-01

    This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

  1. Quantum-Mechanical Calculations on Molecular Substructures Involved in Nanosystems

    Directory of Open Access Journals (Sweden)

    Beata Szefler

    2014-09-01

    Full Text Available In this review article, four ideas are discussed: (a aromaticity of fullerenes patched with flowers of 6-and 8-membered rings, optimized at the HF and DFT levels of theory, in terms of HOMA and NICS criteria; (b polybenzene networks, from construction to energetic and vibrational spectra computations; (c quantum-mechanical calculations on the repeat units of various P-type crystal networks and (d construction and stability evaluation, at DFTB level of theory, of some exotic allotropes of diamond D5, involved in hyper-graphenes. The overall conclusion was that several of the yet hypothetical molecular nanostructures herein described are serious candidates to the status of real molecules.

  2. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

    2008-07-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  3. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    Shwageraus, E.; Fridman, E.

    2008-01-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  4. Cluster monte carlo method for nuclear criticality safety calculation

    International Nuclear Information System (INIS)

    Pei Lucheng

    1984-01-01

    One of the most important applications of the Monte Carlo method is the calculation of the nuclear criticality safety. The fair source game problem was presented at almost the same time as the Monte Carlo method was applied to calculating the nuclear criticality safety. The source iteration cost may be reduced as much as possible or no need for any source iteration. This kind of problems all belongs to the fair source game prolems, among which, the optimal source game is without any source iteration. Although the single neutron Monte Carlo method solved the problem without the source iteration, there is still quite an apparent shortcoming in it, that is, it solves the problem without the source iteration only in the asymptotic sense. In this work, a new Monte Carlo method called the cluster Monte Carlo method is given to solve the problem further

  5. Development of Audit Calculation Methodology for RIA Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joosuk; Kim, Gwanyoung; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    The interim criteria contain more stringent limits than previous ones. For example, pellet-to-cladding mechanical interaction(PCMI) was introduced as a new failure criteria. And both short-term (e.g. fuel-to coolant interaction, rod burst) and long-term(e.g., fuel rod ballooning, flow blockage) phenomena should be addressed for core coolability assurance. For dose calculations, transient-induced fission gas release has to be accounted additionally. Traditionally, the approved RIA analysis methodologies for licensing application are developed based on conservative approach. But newly introduced safety criteria tend to reduce the margins to the criteria. Thereby, licensees are trying to improve the margins by utilizing a less conservative approach. In this situation, to cope with this trend, a new audit calculation methodology needs to be developed. In this paper, the new methodology, which is currently under developing in KINS, was introduced. For the development of audit calculation methodology of RIA safety analysis based on the realistic evaluation approach, preliminary calculation by utilizing the best estimate code has been done on the initial core of APR1400. Followings are main conclusions. - With the assumption of single full-strength control rod ejection in HZP condition, rod failure due to PCMI is not predicted. - And coolability can be assured in view of entalphy and fuel melting. - But, rod failure due to DNBR is expected, and there is possibility of fuel failure at the rated power conditions also.

  6. Safety instruction for execution tasks involving ionizing radiations

    International Nuclear Information System (INIS)

    Fonseca, G.

    1985-01-01

    Basic directives are presented allow operations with ionizing radiations in industrial areas with high levels of safety. Contractual, technical, operational and administrative criteria are established for the safe performance of x-rays and gamographies and the use of fixed radiation based equipment (indicators of level, density, flow, etc) as well as precautions to be taken during project, procurement, transportation, assembly and maintenance of such equipment. Finally procedures are suggested for emergencies involving radioactive sources. (author)

  7. Neutronic calculation of safety parameters for the RP-0 and RP-10 nuclear reactors

    OpenAIRE

    Lázaro, Gerardo; Deen, James R.; Woodruff, William L.

    2002-01-01

    Theoretical safety calculations were done with proved codes utilized by the staff of the RERTR program in the HEU to LEU core conversions. The studies were designed to evaluate the reactivity coefficients and kinetics parameters of the reactor involved in the evolution of peak power transients by reactivity insertion accidents. It was done to show the trend of these reactivity coefficients as a function of the core size and fuel depletion for RP10 cores. It was useful to get a better underst...

  8. Corrosion calculations report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    2010-12-01

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

  9. Corrosion calculations report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

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

  10. Safety reassessment of the old installations involved in fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, M

    2003-01-01

    Since the early 1990's, CEA (French atomic energy commission) has been preparing a plan for the renovation of a laboratory situated at Cadarache and dedicated to the study of irradiated materials and fuels. The main aim of this renovation was the improvement of the seismic behaviour of the laboratory since it was not built according to the para-seismic rules now in force. The solution chosen, given the different projects studied, the provisional unavailability of the plant and the related costs, was a partial reinforcement of the building in association with a limited plant life time and the reduction of activities in the oldest part of the installation. Another aim of this renovation was a global upgrading of the safety concerning: -) radioactive material containment (upgrade of the first static barrier by reinforcing cell leak-proofing, installation of a second level of very high efficiency filtration at the cell outputs, and separation of cell and general building ventilation networks; -) fire protection (fire sectoring with the isolation of the premises involving safety-important equipment, replacement of the automatic fire detection system, and definition of a new piloting of ventilation in case of fire); -) power cut risks (installation of permanent sources for the power supply of safety-important equipment); and -) earthquake behaviour (addition of reinforced connections between the 3 parts of the building, strengthening of peripheral walls, widening of joints between cells and building, and reinforcement of the foundation of the concrete cells). (A.C.)

  11. Safety reassessment of the old installations involved in fuel cycle

    International Nuclear Information System (INIS)

    Guillard, M.

    2003-01-01

    Since the early 1990's, CEA (French atomic energy commission) has been preparing a plan for the renovation of a laboratory situated at Cadarache and dedicated to the study of irradiated materials and fuels. The main aim of this renovation was the improvement of the seismic behaviour of the laboratory since it was not built according to the para-seismic rules now in force. The solution chosen, given the different projects studied, the provisional unavailability of the plant and the related costs, was a partial reinforcement of the building in association with a limited plant life time and the reduction of activities in the oldest part of the installation. Another aim of this renovation was a global upgrading of the safety concerning: -) radioactive material containment (upgrade of the first static barrier by reinforcing cell leak-proofing, installation of a second level of very high efficiency filtration at the cell outputs, and separation of cell and general building ventilation networks; -) fire protection (fire sectoring with the isolation of the premises involving safety-important equipment, replacement of the automatic fire detection system, and definition of a new piloting of ventilation in case of fire); -) power cut risks (installation of permanent sources for the power supply of safety-important equipment); and -) earthquake behaviour (addition of reinforced connections between the 3 parts of the building, strengthening of peripheral walls, widening of joints between cells and building, and reinforcement of the foundation of the concrete cells). (A.C.)

  12. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

  13. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    Chen, S.Y.; MacDonald, R.; MacFarlane, D.

    1983-01-01

    Safety issues for the two general types of reactors, i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl/sub x/) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with HEU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods (approx. 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. Results of transient calculations performed with existing computer codes, most suited for each type of reactor, are presented

  14. Nuclear criticality safety calculational analysis for small-diameter containers

    International Nuclear Information System (INIS)

    LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.

    1995-11-01

    This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can (open-quotes F-canclose quotes); (4) 5.25-inch-diameter by 15-inch-high steel can (open-quotes Z-canclose quotes); and (5) 5.0-inch-diameter by 9-inch-high polybottle (open-quotes CO-4close quotes). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO 2 F 2 +H 2 O and UF 4 +oil materials at 100% and 10% enrichments and U 3 O 8 , and H 2 O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k eff + 2σ < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant

  15. Calculating externalities from damages in occupational health and safety

    Energy Technology Data Exchange (ETDEWEB)

    Burtraw, D; Shefftz, J

    1994-07-01

    This paper surveys the theoretical basis for the possibility that coal miner occupational health and safety damages are not adequately internalized into the production cost of mining coal and thereby impose an external cost on society.

  16. Calculating externalities from damages in occupational health and safety

    International Nuclear Information System (INIS)

    Burtraw, D.; Shefftz, J.

    1994-01-01

    This paper surveys the theoretical basis for the possibility that coal miner occupational health and safety damages are not adequately internalized into the production cost of mining coal and thereby impose an external cost on society

  17. New engineering safety factors for Loviisa NPP core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuopanportti, Jaakko; Saarinen, Simo; Lahtinen, Tuukka; Ekstroem, Karoliina [Fortum Power and Heat Ltd., Fortum (Finland)

    2017-09-15

    In Loviisa NPP, there are two limiting thermal margins called the enthalpy rise margin and the linear heat rate margin that are monitored during normal operation. Engineering safety factors are applied in determination of both of these factors. The factors take into account the effect of various manufacturing tolerances, impact of the irradiation and simulation uncertainties on the local heat rate and on the enthalpy of the coolant. The engineering factors were re-evaluated during 2015 and the factors were approved by the Finnish radiation and nuclear safety authority in 2016. The re-evaluation was performed by considering all of the identified phenomena that affect the local heat rate or the enthalpy of the coolant. This paper summarizes the work that was performed during the re-evaluation of the engineering safety factors and presents the results for each uncertainty component. The new engineering safety factors are 1.115 for the linear heat rate and 1.100 for the enthalpy rise margin when the old factors were 1.12 and 1.16, respectively. The new factors improve the fuel economy by about 1%.

  18. Transport company safety climate - the impact on truck driver behaviour and crash involvement

    OpenAIRE

    Sullman, Mark J. M.; Stephens A. N.; Pajo K.

    2017-01-01

    Objective: The present study investigated the relationships between safety climate and driving behavior and crash involvement. Methods: A total of 339 company-employed truck drivers completed a questionnaire that measured their perceptions of safety climate, crash record, speed choice, and aberrant driving behaviors (errors, lapses, and violations). Results: Although there was no direct relationship between the drivers' perceptions of safety climate and crash involvement, safety clima...

  19. Safety margins and retrofit. The technical calculation perspective; Sicherheitsmargen durch Nachruestung aus Sicht der technischen Berechnung

    Energy Technology Data Exchange (ETDEWEB)

    Daichendt, Matthias [Kraftanlagen Heidelberg GmbH, Heidelberg (Germany). Systemtechnik - Technische Berechnungen

    2016-01-15

    Safety margins are an essential factor of the safety philosophy for nuclear power plants. They support to cover future requirements even today. The basic safety concept is one key topic as also aspects of process engineering, the dimensioning and mechanical analysis of systems and ageing management. Calculations with today's capabilities are an integral part of the determination of safety margins. They can be used to analyse and to assess retrofit measures.

  20. Merger of Nuclear Data with Criticality Safety Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Derrien, H.; Larson, N.M.; Leal, L.C.

    1999-09-20

    In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes � measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations � which in the past have been treated independently.

  1. Merger of Nuclear Data with Criticality Safety Calculations

    International Nuclear Information System (INIS)

    Derrien, H.; Larson, N.M.; Leal, L.C.

    1999-01-01

    In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations which in the past have been treated independently

  2. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

  3. Quality plan for criticality safety calculations at Rocky Flats

    International Nuclear Information System (INIS)

    Pecora, D.

    1978-01-01

    The text of the plan is given, and some of the guidelines followed in writing it are discussed to aid others who may be faced with the same task. The plan is divided into four sections. The Introduction describes the general functions and purpose of the calculational program. The second section, Activities and Responsibilities, lists specific tasks and their purposes and assigns responsibility for performance. The third section references relevant documentation (e.g., ANSI standards), and the final section describes quality plans for specific functions

  4. Neutron flux shape effects in large fast reactor safety calculations

    International Nuclear Information System (INIS)

    Galati, A.; Loizzo, P.; Musco, A.

    1978-01-01

    Three classes of accidents in a large fast reactor were studied by the two-dimensional core dynamics code NADYP-2. A Modified version of the code, including a point kinetics module, allowed comparison between 2D and 0D power, reactivity and temperature histories. A strong shape effect was evidenced by these calculations in the boiling phase of LOF accidents as well as in the accident generated by control rod removal. Some future possibilities of by passing the consequences of this effect are indicated

  5. Institutions involved in food Safety: World Health Organization (WHO)

    DEFF Research Database (Denmark)

    Schlundt, Jørgen

    2014-01-01

    The World Health Organization (WHO) has been a leading intergovernmental organization in the effort to prevent diseases related to food and improve global food safety and security. These efforts have been focused on the provision of independent scientific advice on foodborne risks, the development...... the focus on simple and efficient messaging toward preventing food risks through a better understanding of good food preparation practices in all sectors....

  6. Search for a transport method for the calculation of the PWR control and safety clusters

    International Nuclear Information System (INIS)

    Bruna, G.B.; Van Frank, C.; Vergain, M.L.; Chauvin, J.P.; Palmiotti, G.; Nobile, M.

    1990-01-01

    The project studies of power reactors rely mainly on diffusion calculations, but transport ones are often needed for assessing fine effects, intimately linked to geometry and spectrum heterogeneities. Accurate transport computations are necessary, in particular, for shielded cross section generation, and when homogenization and dishomogenization processes are involved. The transport codes, generally, offer the user a variety of computational options, related to different approximation levels. In every case, it is obviously desirable to be able to choose the reliable degree of approximation to be accepted in any particular computational circumstance of the project. The search for such adapted procedures is to be made on the basis of critical experiments. In our studies, this task was made possible by the availability of suitable results of the CAMELEON critical experiment, carried on in the EOLE facility at CEA's Center of Cadarache. In this paper, we summarize some of the work in progress at FRAMATOME on the definition of an assembly based transport calculation scheme to be used for PWR control and safety cluster computations. Two main items, devoted to the search of the optimum computational procedures, are presented here: - a parametrical study on computational options, made in an infinite medium assembly geometry, - a series of comparisons between calculated and experimental values of pin power distribution

  7. Patient Involvement in Patient Safety: A Qualitative Study of Nursing Staff and Patient Perceptions.

    Science.gov (United States)

    Bishop, Andrea C; Macdonald, Marilyn

    2017-06-01

    The risk associated with receiving health care has called for an increased focus on the role of patients in helping to improve safety. Recent research has highlighted that patient involvement in patient safety practices may be influenced by patient perceptions of patient safety practices and the perceptions of their health care providers. The objective of this research was to describe patient involvement in patient safety practices by exploring patient and nursing staff perceptions of safety. Qualitative focus groups were conducted with a convenience sample of nursing staff and patients who had previously completed a patient safety survey in 2 tertiary hospital sites in Eastern Canada. Six focus groups (June 2011 to January 2012) were conducted and analyzed using inductive thematic analysis. Four themes were identified: (1) wanting control, (2) feeling connected, (3) encountering roadblocks, and (4) sharing responsibility for safety. Both patient and nursing staff participants highlighted the importance of building a personal connection as a precursor to ensuring that patients are involved in their care and safety. However, perceptions of provider stress and nursing staff workload often reduced the ability of the nursing staff and patient participants to connect with one another and promote involvement. Current strategies aimed at increasing patient awareness of patient safety may not be enough. The findings suggest that providing the context for interaction to occur between nursing staff and patients as well as targeted interventions aimed at increasing patient control may be needed to ensure patient involvement in patient safety.

  8. Slope Safety Calculation With A Non-Linear Mohr Criterion Using Finite Element Method

    DEFF Research Database (Denmark)

    Clausen, Johan; Damkilde, Lars

    2005-01-01

    Safety factors for soil slopes are calculated using a non-linear Mohr envelope. The often used linear Mohr-Coulomb envelope tends to overestimate the safety as the material parameters are usually determined at much higher stress levels, than those present at slope failure. Experimental data...

  9. Calculation of partial derivatives of thermophysical properties of sodium for safety analysis

    International Nuclear Information System (INIS)

    Shan Jianqiang; Qiu Suizhang; Zhu Jizhou; Zhang Guiqin

    1997-01-01

    According to the characters of safety analysis of LMFBR, the partial derivatives formula of some special thermophysical properties of sodium, including single-and two-phase properties, are calculated based on the basic Maxwell equations, and on the formulae of basic thermophysical properties of sodium which were verified abroad. The present study can provide theoretical base for safety analysis of LMFBR

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

    International Nuclear Information System (INIS)

    Van Keuren, J.C.

    1997-01-01

    Accident consequence analyses have been performed for Project W-058, the Replacement Cross Site Transfer System. using the assumption and analysis techniques developed for the Tank Remediation Waste system Basis for Interim Operation. most potential accident involving the FISTS are bounded by the TWRS BIO analysis. However, the spray leak and pool leak scenarios require revised analyses since the RCSTS design utilizes larger diameter pipe and higher pressures than those analyzed in the TWRS BIO. Also the volume of diversion box and vent station are larger than that assumed for the valve pits in the TWRS BIO, which effects results of sprays or spills into the pits. the revised analysis for the spray leak is presented in Section 2, for the above ground spill in Section 3, for the presented in Section 2, for the above ground spill in Section 3, for the subsurface spill forming a pool in Section 4, and for the subsurface pool remaining subsurface in Section 5. The conclusion from these sections are summarized below

  11. Trend analysis of incidents involving setpoint drift in safety or safety/relief valves at U.S. LWRs

    International Nuclear Information System (INIS)

    Watanabe, Norio

    2008-01-01

    Since the beginning of the 1980's, in the United States, there have been many licensee event reports (LERs) involving setpoint drift in safety or safety/relief valves. The United States Nuclear Regulatory Commission (NRC) has issued a lot of generic communications on this issue and the industry has made efforts to resolve the issue. However, the NRC staff recently highlighted that over 70 LERs involved instances where safety or safety/relief valves failed to meet the allowed setpoint tolerance from 2001 through August 2006. In the present study, we analyzed the U.S. experience with setpoint drift in safety/relief valves (SRVs) at BWRs, pressurizer safety valves (PSVs), and main steam safety valves (MSSVs) at PWRs by reviewing approximately 90 LERs from 2000 to 2006 and examined the trend focusing on causes and setpoint deviation ranges. This study indicates that for SRVs and MSSVs, disc-seat bonding is a dominant cause of the setpoint drifting high and has a tendency to result in a relatively large deviation of the setpoint. This means that disc-seat bonding might be a safety concern from the view point of overpressure protection. For PSVs, the deviation of setpoints is generally small, although its causes are not specified in many instances. (author)

  12. Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Williams, L.R.

    1994-07-01

    This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

  13. THE NATIONAL AUTHORITY FOR ANIMAL HEALTH AND FOOD SAFETY, THE MAIN BODY INVOLVED IN FOOD SAFETY IN ROMANIA

    Directory of Open Access Journals (Sweden)

    PETRUTA-ELENA ISPAS

    2012-05-01

    Full Text Available This paper is intended to present the role, functions and responsibilities of the National Authority for Animal Health and Food Safety as the main body involved in food safety in Romania. It will be also exposed the Regulation 178/2002 of the European Parliament and the Council, the general food ”law” in Europe, and Law 150/2004, which transposed into Romanian legislation Regulation 178/2002.

  14. Method of calculating the safety factor profile on the HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhang Xianmei; Lu Yuancheng; Wan Baonian

    2001-01-01

    A method of calculating the safety factor profile on the HT-7 tokamak has been described. It is derived from Maxwell's equations, among which the authors mainly use two of them: one is the magnetic field diffusion equation, and the other is Ampere's Law. This method can be also used to evaluate the safety factor on other devices with a circular cross sections. It is helpful to the study of the plasma MHD behavior on the HT-7 tokamak

  15. Alterations in the evaporation and discharge calculations for safety and relief valves in the Almod pressurizer

    International Nuclear Information System (INIS)

    Madeira, A.A.

    1986-01-01

    Models to estimate bubble rise velocity for evaporation, and critical mass flow for pressurizer relief and safety valves discharge calculation were implemented in ALMOD, a digital code developed to perform primary loop simulation of a PWR type during operational transients or accidents without loss of coolant. These models can be utilized alternatively, depending on the requirements for the analyzed transient condition. (Author) [pt

  16. Culture influence and predictors for behavioral involvement in patient safety among hospital nurses in Taiwan.

    Science.gov (United States)

    Chiang, Hui-Ying; Lin, Shu-Yuan; Hsiao, Ya-Chu; Chang, Yuanmay

    2012-01-01

    This study explored the effects of incident reporting culture and willingness of incident reporting on behavioral involvement in patient safety (BIPS) by surveying 1049 hospital nurses in Taiwan. The highest areas of BIPS were handoff communication and discussion on error prevention. Yet, sharing information about human factors toward safety awareness was less frequent. Results indicated that the reporting culture, willingness to report, tenure of work, and reporting rate contributed positively to BIPS.

  17. : Principles of safety measures of sports events organizers without the involvement of police

    OpenAIRE

    Buchalová, Kateřina

    2013-01-01

    Title: Principles of safety measures of sports events organizers without the involvement of police Objectives: The aim of this thesis is a description of security measures at sporting events organizers. Methods: The thesis theoretical style is focused on searching for available sources of study and research, and writing their summary comparing safety measures of the organizers. Results: This work describes the activities of the organizers of sports events and precautions that must be provided...

  18. A comparative analysis between France and Japan on local governments' involvement in nuclear safety governance

    International Nuclear Information System (INIS)

    Sugawara, Shin-etsu; Shiroyama, Hideaki

    2011-01-01

    This paper shows a comparative analysis between France and Japan on the way of the local governments' involvement in nuclear safety governance through some interviews. In France, a law came into force that requires related local governments to establish 'Commision Locale d'Information' (CLI), which means the local governments officially involve in nuclear regulatory activity. Meanwhile, in Japan, related local governments substantially involve in the operation of nuclear facilities through the 'safety agreements' in spite of the lack of legal authority. As a result of comparative analysis, we can point out some institutional input from French cases as follows: to clarify the local governments' roles in the nuclear regulation system, to establish the official channels of communication among nuclear utilities, national regulatory authorities and local governments, and to stipulate explicitly the transparency as a purpose of safety regulation. (author)

  19. Public involvement in environmental, safety and health issues at the DOE Nuclear Weapons Complex

    International Nuclear Information System (INIS)

    Taylor, Laura L.; Morgan, Robert P.

    1992-01-01

    The state of public involvement in environmental, safety, and health issues at the DOE Nuclear Weapons Complex is assessed through identification of existing opportunities for public involvement and through interviews with representatives of ten local citizen groups active in these issues at weapons facilities in their communities. A framework for analyzing existing means of public involvement is developed. On the whole, opportunities for public involvement are inadequate. Provisions for public involvement are lacking in several key stages of the decision-making process. Consequently, adversarial means of public involvement have generally been more effective than cooperative means in motivating change in the Weapons Complex. Citizen advisory boards, both on the local and national level, may provide a means of improving public involvement in Weapons Complex issues. (author)

  20. Establishing credibility in the environmental models used for safety and licensing calculations in the nuclear industry

    International Nuclear Information System (INIS)

    Davis, P.A.

    1997-01-01

    Models that simulate the transport and behaviour of radionuclides in the environment are used extensively in the nuclear industry for safety and licensing purposes. They are needed to calculate derived release limits for new and operating facilities, to estimate consequences following hypothetical accidents and to help manage a real emergency. But predictions generated for these purposes are essentially meaningless unless they are accompanied by a quantitative estimate of the confidence that can be placed in them. For example, in an emergency where there has been an accidental release of radioactivity to the atmosphere, decisions based on a validated model with small uncertainties would likely be very different from those based on an untested model, or on one with large uncertainties. This paper begins with a discussion of some general methods for establishing the credibility of model predictions. The focus will be on environmental transport models but the principles apply to models of all kinds. Establishing the credibility of a model is not a trivial task, It involves a number of tasks including face validation, verification, experimental validation and sensitivity and uncertainty analyses. The remainder of the paper will present quantitative results relating to the credibility of environmental transport models. Model formation, choice of parameter values and the influence of the user will all be discussed as sources of uncertainty in predictions. The magnitude of uncertainties that must be expected in various applications of the models will be presented. The examples used throughout the paper are drawn largely from recent work carried out in BIOMOVS and VAMP. (DM)

  1. Safety reloaded: lean operations and high involvement work practices for sustainable workplaces

    OpenAIRE

    Camuffo, Arnaldo; De Stefano, Federica; Paolino, Chiara

    2017-01-01

    Starting from the recent quest to investigate the human side of organizational sustainability, this study applies a variety of regression analyses to investigate the effects of Lean Operations, High Involvement Work Practices, and management behaviors on occupational safety. It tests and finds support for the hypotheses that Lean Production systems, High Involvement Work Practices, and two specific management behaviors—workers’ capability development (coaching and teaching of workers) and emp...

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

    Directory of Open Access Journals (Sweden)

    Herrero J.J.

    2017-01-01

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

  3. Planning and Preparing for Emergency Response to Transport Accidents Involving Radioactive Material. Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    This Safety Guide provides guidance on various aspects of emergency planning and preparedness for dealing effectively and safely with transport accidents involving radioactive material, including the assignment of responsibilities. It reflects the requirements specified in Safety Standards Series No. TS-R-1, Regulations for the Safe Transport of Radioactive Material, and those of Safety Series No. 115, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. Contents: 1. Introduction; 2. Framework for planning and preparing for response to accidents in the transport of radioactive material; 3. Responsibilities for planning and preparing for response to accidents in the transport of radioactive material; 4. Planning for response to accidents in the transport of radioactive material; 5. Preparing for response to accidents in the transport of radioactive material; Appendix I: Features of the transport regulations influencing emergency response to transport accidents; Appendix II: Preliminary emergency response reference matrix; Appendix III: Guide to suitable instrumentation; Appendix IV: Overview of emergency management for a transport accident involving radioactive material; Appendix V: Examples of response to transport accidents; Appendix VI: Example equipment kit for a radiation protection team; Annex I: Example of guidance on emergency response to carriers; Annex II: Emergency response guide.

  4. Safety analysis calculations for a mixed and full FLIP core in a TRIGA Mark II

    International Nuclear Information System (INIS)

    Ringle, John C.; Hornyik, K.; Robinson, A.H.; Anderson, T.V.; Johnson, A.G.

    1976-01-01

    The Oregon State TRIGA Reactor will be reloading with FLIP fuel in August 1976. As we are the first Mark II TRIGA with a circular grid pattern and graphite reflector to utilize FLIP fuel, the safety analysis calculations performed at other facilities using FLIP were only of limited use to us. A multigroup, multiregion, one-dimensional diffusion theory code was used to calculate power densities in six different operational cores - mixed to full FLIP. Pulsing characteristics were obtained from a computer code based on point kinetics, with adiabatic heating of the fuel, linear temperature dependence of the specific heat, and prompt fuel temperature feedback coefficient. The results of all pertinent calculations will be presented. (author)

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

    International Nuclear Information System (INIS)

    Pepin, G.; Sallaberry, C.

    2003-01-01

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

  6. Calculation of the state of safety (SOS) for lithium ion batteries

    Science.gov (United States)

    Cabrera-Castillo, Eliud; Niedermeier, Florian; Jossen, Andreas

    2016-08-01

    As lithium ion batteries are adopted in electric vehicles and stationary storage applications, the higher number of cells and greater energy densities increases the risks of possible catastrophic events. This paper shows a definition and method to calculate the state of safety of an energy storage system based on the concept that safety is inversely proportional to the concept of abuse. As the latter increases, the former decreases to zero. Previous descriptions in the literature are qualitative in nature but don't provide a numerical quantification of the safety of a storage system. In the case of battery testing standards, they only define pass or fail criteria. The proposed state uses the same range as other commonly used state quantities like the SOC, SOH, and SOF, taking values between 0, completely unsafe, and 1, completely safe. The developed function combines the effects of an arbitrary number of subfunctions, each of which describes a particular case of abuse, in one or more variables such as voltage, temperature, or mechanical deformation, which can be detected by sensors or estimated by other techniques. The state of safety definition can be made more general by adding new subfunctions, or by refining the existing ones.

  7. Involving patients in patient safety programmes: A scoping review and consensus procedure by the LINNEAUS collaboration on patient safety in primary care

    NARCIS (Netherlands)

    Trier, H.; Valderas, J.M.; Wensing, M.; Martin, H.M.; Egebart, J.

    2015-01-01

    BACKGROUND: Patient involvement has only recently received attention as a potentially useful approach to patient safety in primary care. OBJECTIVE: To summarize work conducted on a scoping review of interventions focussing on patient involvement for patient safety; to develop consensus-based

  8. International report to validate criticality safety calculations for fissile material transport

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1984-01-01

    During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

  9. Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  10. Perturbative methods for sensitivity calculation in safety problems of nuclear reactors: state-of-the-art

    International Nuclear Information System (INIS)

    Lima, Fernando R.A.; Lira, Carlos A.B.O.; Gandini, Augusto

    1995-01-01

    During the last two decades perturbative methods became an efficient tool to perform sensitivity analysis in nuclear reactor safety problems. In this paper, a comparative study taking into account perturbation formalisms (Diferential and Matricial Mthods and generalized Perturbation Theory - GPT) is considered. Then a few number of applications are described to analyze the sensitivity of some functions relavant to thermal hydraulics designs or safety analysis of nuclear reactor cores and steam generators. The behaviours of the nuclear reactor cores and steam generators are simulated, respectively, by the COBRA-IV-I and GEVAP codes. Results of sensitivity calculations have shown a good agreement when compared to those obtained directly by using the mentioned codes. So, a significative computational time safe can be obtained with perturbative methods performing sensitivity analysis in nuclear power plants. (author). 25 refs., 5 tabs

  11. Planning and preparing for emergency response to transport accidents involving radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of this Safety Guide is to provide guidance to the public authorities and others (including consignors, carriers and emergency response authorities) who are responsible for developing and establishing emergency arrangements for dealing effectively and safely with transport accidents involving radioactive material. It may assist those concerned with establishing the capability to respond to such transport emergencies. It provides guidance for those Member States whose involvement with radioactive material is just beginning. It also provides guidance for those Member States that have already developed their radioactive material industries and the attendant emergency plans but that may need to review and improve these plans

  12. Slope Safety Factor Calculations With Non-Linear Yield Criterion Using Finite Elements

    DEFF Research Database (Denmark)

    Clausen, Johan; Damkilde, Lars

    2006-01-01

    The factor of safety for a slope is calculated with the finite element method using a non-linear yield criterion of the Hoek-Brown type. The parameters of the Hoek-Brown criterion are found from triaxial test data. Parameters of the linear Mohr-Coulomb criterion are calibrated to the same triaxial...... are carried out at much higher stress levels than present in a slope failure, this leads to the conclusion that the use of the non-linear criterion leads to a safer slope design...

  13. Involving patients in patient safety programmes: A scoping review and consensus procedure by the LINNEAUS collaboration on patient safety in primary care.

    Science.gov (United States)

    Trier, Hans; Valderas, Jose M; Wensing, Michel; Martin, Helle Max; Egebart, Jonas

    2015-09-01

    Patient involvement has only recently received attention as a potentially useful approach to patient safety in primary care. To summarize work conducted on a scoping review of interventions focussing on patient involvement for patient safety; to develop consensus-based recommendations in this area. Scoping review of the literature 2006-2011 about methods and effects of involving patients in patient safety in primary care identified evidence for previous experiences of patient involvement in patient safety. This information was fed back to an expert panel for the development of recommendations for healthcare professionals and policy makers. The scoping review identified only weak evidence in support of the effectiveness of patient involvement. Identified barriers included a number of patient factors but also the healthcare workers' attitudes, abilities and lack of training. The expert panel recommended the integration of patient safety in the educational curricula for healthcare professionals, and expected a commitment from professionals to act as first movers by inviting and encouraging the patients to take an active role. The panel proposed a checklist to be used by primary care clinicians at the point of care for promoting patient involvement. There is only weak evidence on the effectiveness of patient involvement in patient safety. The recommendations of the panel can inform future policy and practice on patient involvement in safety in primary care.

  14. Patient involvement in blood transfusion safety: patients' and healthcare professionals' perspective.

    Science.gov (United States)

    Davis, R; Murphy, M F; Sud, A; Noel, S; Moss, R; Asgheddi, M; Abdur-Rahman, I; Vincent, C

    2012-08-01

    Blood transfusion is one of the major areas where serious clinical consequences, even death, related to patient misidentification can occur. In the UK, healthcare professional compliance with pre-transfusion checking procedures which help to prevent misidentification errors is poor. Involving patients at a number of stages in the transfusion pathway could help prevent the occurrence of these incidents. To investigate patients' willingness to be involved and healthcare professionals' willingness to support patient involvement in pre-transfusion checking behaviours. A cross-sectional design was employed assessing willingness to participate in pre-transfusion checking behaviours (patient survey) and willingness to support patient involvement (healthcare professional survey) on a scale of 1-7. One hundred and ten patients who had received a transfusion aged between 18 and 93 (60 male) and 123 healthcare professionals (doctors, nurses and midwives) involved in giving blood transfusions to patients. Mean scores for patients' willingness to participate in safety-relevant transfusion behaviours and healthcare professionals' willingness to support patient involvement ranged from 4.96-6.27 to 4.53-6.66, respectively. Both groups perceived it most acceptable for patients to help prevent errors or omissions relating to their hospital identification wristband. Neither prior experience of receiving a blood transfusion nor professional role of healthcare staff had an effect on attitudes towards patient participation. Overall, both patients and healthcare professionals view patient involvement in transfusion-related behaviours quite favourably and appear in agreement regarding the behaviours patients should adopt an active role in. Further work is needed to determine the effectiveness of this approach to improve transfusion safety. © 2012 The Authors. Transfusion Medicine © 2012 British Blood Transfusion Society.

  15. Activation calculation and environmental safety analysis for fusion experimental breeder (FEB)

    Energy Technology Data Exchange (ETDEWEB)

    Kaiming, Feng [Southwest Inst. of Physics, Leshan, SC (China)

    1996-04-01

    An activation calculation code FDKR and decay chain data library AFDCDLIB are used to calculate the radioactivity, decay heat, dose rate and biological hazard potential (BHP) form activation products, actinides and fission products in a Fusion Experiment Breeder (FEB). The code and library are introduced briefly, and calculation results and decay curves of related hazards after one year operation with 150 MW fusion power are given. The total radioactivity inventory, decay heat and BHP are 5.74 x 10{sup 20} Bq, 8.34 MW and 4.08 x 10{sup 8} km{sup 3} of air, respectively, at shutdown. Results obtained show that the first wall of FEB can meet the nuclear waste disposal criteria for the NRC 10 CFR61 Class C after a few weeks from shutdown. The inventory of important actinides for the fuel reprocessing, such as {sup 232}U and {sup 237}Np were also calculated. It was shown that their concentrations do not excess the limit value of environmental safety required. (9 refs., 4 figs., 9 tabs.).

  16. First example of a high-level correlated calculation of the indirect spin-spin coupling constants involving tellurium

    DEFF Research Database (Denmark)

    Rusakov, Yury Yu; Krivdin, Leonid B.; Østerstrøm, Freja From

    2013-01-01

    This paper documents a very first example of a high-level correlated calculation of spin-spin coupling constants involving tellurium taking into account relativistic effects, vibrational corrections and solvent effects for the medium sized organotellurium molecules. The 125Te-1H spin-spin coupling...... constants of tellurophene and divinyl telluride were calculated at the SOPPA and DFT levels in a good agreement with experiment. A new full-electron basis set av3z-J for tellurium derived from the "relativistic" Dyall's basis set, dyall.av3z, and specifically optimized for the correlated calculations...... of spin-spin coupling constants involving tellurium, was developed. The SOPPA methods show much better performance as compared to 15 those of DFT, if relativistic effects calculated within the ZORA scheme are taken into account. Vibrational and solvent corrections are next to negligible, while...

  17. Involvement of patients with cancer in patient safety: a qualitative study of current practices, potentials and barriers.

    Science.gov (United States)

    Martin, Helle Max; Navne, Laura Emdal; Lipczak, Henriette

    2013-10-01

    Patient involvement in patient safety is widely advocated but knowledge regarding implementation of the concept in clinical practice is sparse. To investigate existing practices for patient involvement in patient safety, and opportunities and barriers for further involvement. A qualitative study of patient safety involvement practices in patient trajectories for prostate, uterine and colorectal cancer in Denmark. Observations from four hospital wards and interviews with 25 patients with cancer, 11 hospital doctors, 10 nurses, four general practitioners and two private practicing gynaecologists were conducted using ethnographic methodology. Patient safety was not a topic of attention for patients or dominant in communication between patients and healthcare professionals. The understanding of patient safety in clinical practice is almost exclusively linked to disease management. Involvement of patients is not systematic, but healthcare professionals and patients express willingness to engage. Invitation and encouragement of patients to become involved could be further systematised and developed. Barriers include limited knowledge of patient safety, of specific patient safety involvement techniques and concern regarding potential negative impact on doctor-patient relationship. Involvement of patients in patient safety must take into account that despite stated openness to the idea of involvement, patients and health professionals may not in practice show immediate concern. Lack of systematic involvement can also be attributed to limited knowledge about how to implement involvement beyond the focus of self-monitoring and compliance and a concern about the consequences of patient involvement for treatment outcomes. To realise the potential of patients' and health professionals' shared openness towards involvement, there is a need for more active facilitation and concrete guidance on how involvement can be practiced by both parties.

  18. Involvement of AVN as TSO in the safety analysis of radioactive waste disposal

    International Nuclear Information System (INIS)

    Gelder, P. de; Nys, V.; Smidts, O.; Boeck, B. de

    2004-01-01

    In 1998, ONDRAF/NIRAS, the agency responsible for radioactive waste management in Belgium, was requested by the government to involve the nuclear safety authorities in its activities of safety evaluation of site-specific waste disposal options (deep or surface disposal) for the short-lived low-level waste. A working group was created in which ONDRAF/NIRAS, FANC (the Federal Agency for Nuclear Control) and AVN discuss different aspects of the ONDRAF/NIRAS program concerning the long-term management of short-lived low-level radioactive waste disposal. It includes also the review of technical safety assessments performed by ONDRAF/NIRAS or by contractors for ONDRAF/NIRAS. The involvement of AVN (the Belgian TSO) in the pre-project phase appears to be positive for all partners. Indeed, all felt the need for an independent actor, with a strong technical basis. Through this presentation, the experience and the topics discussed since 1998 will be developed. Mainly, the presentation will focus on the approach followed to develop competency in the radioactive waste field, on the discussions about the development of a regulatory framework adapted to final disposal of low-level radioactive waste, and on the technical regulatory positions developed so far. Also the experience related to the interaction with local stakeholders will be described. (orig.)

  19. Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

    International Nuclear Information System (INIS)

    Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

    2013-01-01

    The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

  20. Analysis of a calculation method for the determination of the value of safety or control bars

    International Nuclear Information System (INIS)

    Aguilar H, F.; Torres A, C.; Filio L, C.

    1982-09-01

    Due to the control or safety bars in a nuclear reactor are constituted by strongly absorbent materials, the Diffusion Theory like tool for the calculation of bar values is not directly applicable, should it use the Transport Theory. However the speed and economy of the Diffusion codes for the reactors calculation, those make attractiveness and by this reason its are used in the determination of characteristic parameters and even in the determination of bar values, not without before to make some theoretical developments that allow to make applicable this theory. The application of the Diffusion Theory in strongly absorbent media is based on the use of some effective cross sections distinct from the real ones obtained when imposing the reason that among the flow and it gradient in the external surface of such media (control element in general, bar type or flagstone) be similar to the one obtained using Transport Theory in all the control region (multiplicative and absorbent media) with those real cross sections. The effective cross sections were obtained of the Leopard-NUMICE cell code which has incorporate the respective calculation theory of effective cross sections. Later these constants its were used in the bidimensional diffusion code Exterminator-II, simulating in it, the distribution of safety or control bars. From the cell code its were also obtained the respective constants of the homogeneous fuel cell. The results as soon as those obtained bar values of the diffusion code, its were compared with some experimental results obtained in the Rφ Swedish reactor of natural uranium and heavy water. In this work an analysis of the bar value of one of them, trying to determine the applicability of the method is made. (Author)

  1. Importance of LWR best-estimate safety calculations for analysis of Fukushima-like accidents

    International Nuclear Information System (INIS)

    Sanchez Espinoza, V.; Ivanov, K.

    2011-01-01

    The safety assessment of nuclear power plants relies heavily on numerical simulations, which must include the most important physical models that are representative for the reactor type of interest. The current trends in nuclear power generation and regulation are to perform safety studies by 'best-estimate' codes that allow a realistic modeling of nuclear and thermal-hydraulic processes of the reactor core and the entire plant behavior including control and protection functions. Realistic methods are referred to as 'best-estimate' calculations, implying that they use a set of data, correlations, and methods designed to represent the phenomena, using the best available techniques. The application of best-estimate methodologies in the licensing process requires the quantification of the embedded uncertainties of the used codes. In this field many international initiatives are underway under the umbrella of the OECD such as the Light Water Reactor Uncertainty Analysis in Modeling benchmark, Oskarshamn 2 Boiling Water Reactor (BWR) Stability benchmark, Kalinin-3 VVER-1000 benchmark, etc. that underlies the importance of these issues. The Fukushima accident has shown the importance of the knowledge of the initial phase of the accident regarding the state of the core, in-vessel structures, and containment as well as the amount of fissile material inventories that potentially can be released if the safety barriers fail. For the development of mitigation and prevention measures modeling of the sequence of the events along with understanding of the key physical phenomena driving the accident progression is important. The paper presents the best-estimate coupled methodologies implemented, validated and applied at the Karlsruhe Institute Technology (KIT) for both types of LWRs - Pressurized Water Reactors (PWRs) and BWRs. Example are given with a BWR steady state and transient simulations along with corresponding uncertainty quantification. The on-going development of high

  2. Institutional Oversight of Occupational Health and Safety for Research Programs Involving Biohazards.

    Science.gov (United States)

    Dyson, Melissa C; Carpenter, Calvin B; Colby, Lesley A

    2017-06-01

    Research with hazardous biologic materials (biohazards) is essential to the progress of medicine and science. The field of microbiology has rapidly advanced over the years, partially due to the development of new scientific methods such as recombinant DNA technology, synthetic biology, viral vectors, and the use of genetically modified animals. This research poses a potential risk to personnel as well as the public and the environment. Institutions must have appropriate oversight and take appropriate steps to mitigate the risks of working with these biologic hazards. This article will review responsibilities for institutional oversight of occupational health and safety for research involving biologic hazards.

  3. Experimental validation of calculated capture rate for nucleus involved in fuel cycle

    International Nuclear Information System (INIS)

    Benslimane-Bouland, A.

    1997-01-01

    This work has been realized in the framework of the estimation of actinides and fission products nuclear data for the today and future reactors. The first part presents the existing integral experiments for the calculated capture rate and the methods used in the design of reactor cores calculation formulary. The second part is devoted to the interpretation of three specific irradiation experiments which allow the evaluation of the today knowledge on studied data and their associated uncertainties. The last part presents a synthesis of results and the statistical methods used for the adjustment of data bases. This work shows that, in spite of the reactors Physics progresses on the knowledge of uranium and plutonium capture cross sections, uncertainties remain for minor actinides. (A.L.B.)

  4. Involving High School Students in Computational Physics University Research: Theory Calculations of Toluene Adsorbed on Graphene.

    Science.gov (United States)

    Ericsson, Jonas; Husmark, Teodor; Mathiesen, Christoffer; Sepahvand, Benjamin; Borck, Øyvind; Gunnarsson, Linda; Lydmark, Pär; Schröder, Elsebeth

    2016-01-01

    To increase public awareness of theoretical materials physics, a small group of high school students is invited to participate actively in a current research projects at Chalmers University of Technology. The Chalmers research group explores methods for filtrating hazardous and otherwise unwanted molecules from drinking water, for example by adsorption in active carbon filters. In this project, the students use graphene as an idealized model for active carbon, and estimate the energy of adsorption of the methylbenzene toluene on graphene with the help of the atomic-scale calculational method density functional theory. In this process the students develop an insight into applied quantum physics, a topic usually not taught at this educational level, and gain some experience with a couple of state-of-the-art calculational tools in materials research.

  5. Nuclear criticality safety calculations for a K-25 site vacuum cleaner

    International Nuclear Information System (INIS)

    Shor, J.T.; Haire, M.J.

    1997-02-01

    A modified Nilfisk model GSJ dry vacuum cleaner is used throughout the K-25 Site to collect dry forms of highly enriched uranium (HEU). When vacuuming, solids are collected in a cyclone-type separator vacuum cleaner body. Calculations were done with the SCALE (KENO V.a) computer code to establish conditions at which a nuclear criticality event might occur if the vacuum cleaner was filled with fissile solution. Conditions evaluated included full (12-in. water) reflection and nominal (1-in. water) reflection, and full (100%) and 20% 235 U enrichment. Validation analyses of SCALE/KENO and the SCALE 27-group cross sections for nuclear criticality safety applications indicate that a calculated k eff + 2σ eff + 2σ ≥ 0.9605 is considered unsafe and may be critical. Critical conditions were calculated to be 70 g U/L for 100% 235 U and full 12-in. water reflection. This corresponds to a minimum critical mass of approximately 1,400 g 235 U for the approximate 20.0-L volume of the vacuum cleaner. The actual volume of the vacuum cleaner is smaller than the modeled volume because some internal materials of construction were assumed to be fissile solution. The model was an overestimate, for conservatism, of fissile solution occupancy. At nominal reflection conditions, the critical concentration in a vacuum cleaner full of UO 2 F 2 solution was calculated to be 100 g 235 U/L, or 2,000 g mass of 100% 235 U. At 20% 235 U for the 20.0-L volume of the vacuum cleaner. At 15% 235 U enrichment and full reflection, critical conditions were not reached at any possible concentration of uranium as a uranyl fluoride solution. At 17.5% 235 U enrichment, criticality was reached at approximately 1,300 g U/L which is beyond saturation at 25 C

  6. Exploration of Important Issues for the Safety of SFR 1 using Performance Assessment Calculations

    International Nuclear Information System (INIS)

    Maul, P.R.; Robinson, P.C.

    2002-06-01

    SKB has produced a revised safety case for the SFR 1 disposal facility for low and intermediate level radioactive wastes at Forsmark: project SAFE. This assessment includes a Performance Assessment (PA) for the long term post-closure safety of the facility. SKI has a responsibility to scrutinise SKB's safety case that is shared with SSI. Quintessa has undertaken a review of SKB's case for the long term safety of SFR 1 to assist SKI's evaluation of SAFE, and this is given in SKI-R--02-61, henceforth referred to as the Quintessa Review. The current report describes the independent PA calculations that provided an input to that review. Since 1999 SKI has been developing a PA capability for SFR 1 using the AMBER software. Two key features of the approach taken have been: To represent the whole system in a single model; and To allow the time-dependency of all key features, events and processes to be represented. These capabilities allow a better understanding of the key features of the system to be obtained for different future evolutions (scenarios). This report presents a summary of the work undertaken to provide SKI with a PA capability for SFR 1 and the calculations undertaken with it. Calculations have been undertaken for radionuclides transported in groundwater and gas, but not for direct intrusion by humans into the wastes. It should be emphasised that the purpose of the Performance Assessment calculations described in this report is not to provide an alternative assessment of potential radiological impacts to that produced by SKB. The aim is to use the models that have been developed to investigate the important features of the system and to help SKI scrutinise the case put to them by SKB. The PA calculations that have been undertaken are by no means comprehensive, and various issues could be investigated further if required. The key issues that have been identified can be summarised as follows: 1. The SFR 1 system has a number of different timescales that can

  7. Shields calculations for teletherapy equipment. Regulatory approach of the National Center of Nuclear Safety

    International Nuclear Information System (INIS)

    Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y.

    2006-01-01

    The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)

  8. Neutron flux calculations for criticality safety analysis using the narrow resonance approximations. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A M [National Center for Nuclear Safety and Radiation Control, NC-NSRC, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    The narrow resonance approximation is applicable for all low-energy resonances and the heaviest nuclides. It is of great importance in neutron calculations, hence, fertile isotopes do not undergo fission at resonance energies. The effect of overestimating the self shielded group averaged cross-section data for a given resonance nuclide can be fairly serious. In the present work, a detailed study, and derivation of the problem of self-shielding are carried-out through the information of Hansen-roach library which is used for criticality safety analysis. The intermediate neutron flux spectrum is analyzed, using the narrow resonance approximation. The resonance self-shielded values of various cross-sections are determined. 4 figs., 3 tabs.

  9. Gas-Induced Water-hammer Loads Calculation for Safety Related Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seungchan; Yoon, Dukjoo [Korea Hydro and Nuclear Power Co., LTd, Daejeon (Korea, Republic of); Lee, Dooyong [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    Of particular interest, gas accumulation can result in system pressure transient in pump discharge piping following a pump start. Consequently, this evolves into a gas-water, a water-hammer event and the accompanying force imbalances on the piping segments can be sufficient to challenge the piping supports and restraint. This paper describes an method performing to the water-hammer loads to determine the maximum loading that would occur in the piping system following the safety injection signal and to evaluate its integrity. For a given gas void volumes in the discharge piping, the result of the calculation shows the maximum loads of 18,894.2psi, which is smaller than the allowable criteria. Also, the maximum peak axial force imbalances acting on the support is 1,720lbf as above.

  10. Gas-Induced Water-hammer Loads Calculation for Safety Related Systems

    International Nuclear Information System (INIS)

    Lee, Seungchan; Yoon, Dukjoo; Lee, Dooyong

    2013-01-01

    Of particular interest, gas accumulation can result in system pressure transient in pump discharge piping following a pump start. Consequently, this evolves into a gas-water, a water-hammer event and the accompanying force imbalances on the piping segments can be sufficient to challenge the piping supports and restraint. This paper describes an method performing to the water-hammer loads to determine the maximum loading that would occur in the piping system following the safety injection signal and to evaluate its integrity. For a given gas void volumes in the discharge piping, the result of the calculation shows the maximum loads of 18,894.2psi, which is smaller than the allowable criteria. Also, the maximum peak axial force imbalances acting on the support is 1,720lbf as above

  11. Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

  12. A pedestal temperature model with self-consistent calculation of safety factor and magnetic shear

    International Nuclear Information System (INIS)

    Onjun, T; Siriburanon, T; Onjun, O

    2008-01-01

    A pedestal model based on theory-motivated models for the pedestal width and the pedestal pressure gradient is developed for the temperature at the top of the H-mode pedestal. The pedestal width model based on magnetic shear and flow shear stabilization is used in this study, where the pedestal pressure gradient is assumed to be limited by first stability of infinite n ballooning mode instability. This pedestal model is implemented in the 1.5D BALDUR integrated predictive modeling code, where the safety factor and magnetic shear are solved self-consistently in both core and pedestal regions. With the self-consistently approach for calculating safety factor and magnetic shear, the effect of bootstrap current can be correctly included in the pedestal model. The pedestal model is used to provide the boundary conditions in the simulations and the Multi-mode core transport model is used to describe the core transport. This new integrated modeling procedure of the BALDUR code is used to predict the temperature and density profiles of 26 H-mode discharges. Simulations are carried out for 13 discharges in the Joint European Torus and 13 discharges in the DIII-D tokamak. The average root-mean-square deviation between experimental data and the predicted profiles of the temperature and the density, normalized by their central values, is found to be about 14%

  13. Reference quantum chemical calculations on RNA base pairs directly involving the 2'-OH group of ribose

    Czech Academy of Sciences Publication Activity Database

    Šponer, Jiří; Zgarbová, M.; Jurečka, Petr; Riley, K.E.; Šponer, Judit E.; Hobza, Pavel

    2009-01-01

    Roč. 5, č. 4 (2009), s. 1166-1179 ISSN 1549-9618 R&D Projects: GA AV ČR(CZ) IAA400040802; GA AV ČR(CZ) IAA400550701; GA MŠk(CZ) LC06030; GA MŠk(CZ) LC512 Institutional research plan: CEZ:AV0Z50040507; CEZ:AV0Z50040702; CEZ:AV0Z40550506 Keywords : RNA * ribose * quantum calculations Subject RIV: BO - Biophysics Impact factor: 4.804, year: 2009

  14. Involving fathers in teaching youth about farm tractor seatbelt safety--a randomized control study.

    Science.gov (United States)

    Jinnah, Hamida Amirali; Stoneman, Zolinda; Rains, Glen

    2014-03-01

    Farm youth continue to experience high rates of injury and deaths as a result of agricultural activities. Farm machinery, especially tractors, is the most common cause of casualties to youth. A Roll-Over Protection Structure (ROPS) along with a fastened seatbelt can prevent almost all injuries and fatalities from tractor overturns. Despite this knowledge, the use of seatbelts by farmers on ROPS tractors remains low. This study treats farm safety as a family issue and builds on the central role of parents as teachers and role models of farm safety for youth. This research study used a longitudinal, repeated-measures, randomized-control design in which youth 10-19 years of age were randomly assigned to either of two intervention groups (parent-led group and staff-led group) or the control group. Fathers in the parent-led group were less likely to operate ROPS tractors without a seatbelt compared with other groups. They were more likely to have communicated with youth about the importance of wearing seatbelts on ROPS tractors. Consequently, youth in the parent-led group were less likely to operate a ROPS tractor without a seatbelt than the control group at post-test. This randomized control trial supports the effectiveness of a home-based, father-led farm safety intervention as a promising strategy for reducing youth as well as father-unsafe behaviors (related to tractor seatbelts) on the farm. This intervention appealed to fathers' strong motivation to practice tractor safety for the sake of their youth. Involving fathers helped change both father as well as youth unsafe tractor-seatbelt behaviors. Copyright © 2014 Society for Adolescent Health and Medicine. Published by Elsevier Inc. All rights reserved.

  15. RELAP5/MOD3 assessment for calculation of safety and relief valve discharge piping hydrodynamic loads

    International Nuclear Information System (INIS)

    Stubbe, E.J.; VanHoenacker, L.; Otero, R.

    1994-02-01

    This report presents an assessment study for the use of the code RELAP 5/MOD3/5M5 in the calculation of transient hydrodynamic loads on safety and relief discharge pipes. Its predecessor, RELAP 5/MOD1, was found adequate for this kind of calculations by EPRI. The hydrodynamic loads are very important for the discharge piping design because of the fast opening of the valves and the presence of liquid in the upstream loop seals. The code results are compared to experimental load measurements performed at the Combustion Engineering Laboratory in Windsor (US). Those measurements were part of the PWR Valve Test Program undertaken by EPRI after the TMI-2 accident. This particular kind of transients challenges the applicability of the following code models: two-phase choked discharge; interphase drag in conditions with large density gradients; heat transfer to metallic structures in fast changing conditions; two-phase flow at abrupt expansions. The code applicability to this kind of transients is investigated. Some sensitivity analyses to different code and model options are performed. Finally, the suitability of the code and some modeling guidelines are discussed

  16. Experimental validation of calculated capture rate for nucleus involved in fuel cycle

    International Nuclear Information System (INIS)

    Benslimane-Bouland, A.

    1997-09-01

    The framework of this study was the evaluation of the nuclear data requirements for Actinides and Fission Products applied to current nuclear reactors as well as future applications. This last item includes extended irradiation campaigns, 100 % Mixed Oxide fuel, transmutation or even incineration. The first part of this study presents different types of integral measurements which are available for capture rate measurements, as well as the methods used for reactor core calculation route design and nuclear data library validation. The second section concerns the analysis of three specific irradiation experiments. The results have shown the extent of the current knowledge on nuclear data as well as the associated uncertainties. The third and last section shows both the coherency between all the results, and the statistical method applied for nuclear data library adjustment. A relevant application of this method has demonstrated that only specifically chosen integral experiments can be of use for the validation of nuclear data libraries. The conclusion is reached that even if co-ordinated efforts between reactor and nuclear physicists have made possible a huge improvement in the knowledge of capture cross sections of the main nuclei such as uranium and plutonium, some improvements are currently necessary for the minor actinides (Np, Am and Cm). Both integral and differential measurements are recommended to improve the knowledge of minor actinide cross sections. As far as integral experiments are concerned, a set of criteria to be followed during the experimental conception have been defined in order to both reduce the number of required calculation approximations, and to increase as much as possible the maximum amount of extracted information. (author)

  17. Calculation of combustible waste fraction (CWF) estimates used in organics safety issue screening

    International Nuclear Information System (INIS)

    Heasler, P.G.; Gao, F.; Toth, J.J.

    1998-08-01

    This report describes how in-tank measurements of moisture (H 2 O) and total organic carbon (TOC) are used to calculate combustible waste fractions (CWF) for 138 of the 149 Hanford single shell tanks. The combustible waste fraction of a tank is defined as that proportion of waste that is capable of burning when exposed to an ignition source. These CWF estimates are used to screen tanks for the organics complexant safety issue. Tanks with a suitably low fraction of combustible waste are classified as safe. The calculations in this report determine the combustible waste fractions in tanks under two different moisture conditions: under current moisture conditions, and after complete dry out. The first fraction is called the wet combustible waste fraction (wet CWF) and the second is called the dry combustible waste fraction (dry CWF). These two fractions are used to screen tanks into three categories: if the wet CWF is too high (above 5%), the tank is categorized as unsafe; if the wet CWF is low but the dry CWF is too high (again, above 5%), the tank is categorized as conditionally safe; finally, if both the wet and dry CWF are low, the tank is categorized as safe. Section 2 describes the data that was required for these calculations. Sections 3 and 4 describe the statistical model and resulting fit for dry combustible waste fractions. Sections 5 and 6 present the statistical model used to estimate wet CWF and the resulting fit. Section 7 describes two tests that were performed on the dry combustible waste fraction ANOVA model to validate it. Finally, Section 8 presents concluding remarks. Two Appendices present results on a tank-by-tank basis

  18. Nuclear installations: if the biotechnologist is involved sooner in the evaluation of design, safety worries are better integrated

    International Nuclear Information System (INIS)

    Charron, S.; Tosello, M.

    1995-01-01

    The institutional background to the safety assessment of nuclear installations is based upon tripartite links between the operator of a complex and hazardous process, the regulatory authorities and their technical support services. The biotechnologists responsible for the human factor side of the safety assessment are better able to deal with this complex situation if they get involved at the very outset of a project: in order to reach a compromise that is more acceptable from the safety standpoint. (authors). 7 refs

  19. An examination of safety reports involving electronic flight bags and portable electronic devices

    Science.gov (United States)

    2014-06-01

    The purpose of this research was to develop a better understanding of safety considerations with the use of Electronic Flight Bags (EFBs) and Portable Electronic Devices (PEDs) by examining safety reports from Aviation Safety Reporting System (ASRS),...

  20. Sophisticated Calculation of the 1oo4-architecture for Safety-related Systems Conforming to IEC61508

    International Nuclear Information System (INIS)

    Hayek, A; Al Bokhaiti, M; Schwarz, M H; Boercsoek, J

    2012-01-01

    With the publication and enforcement of the standard IEC 61508 of safety related systems, recent system architectures have been presented and evaluated. Among a number of techniques and measures to the evaluation of safety integrity level (SIL) for safety-related systems, several measures such as reliability block diagrams and Markov models are used to analyze the probability of failure on demand (PFD) and mean time to failure (MTTF) which conform to IEC 61508. The current paper deals with the quantitative analysis of the novel 1oo4-architecture (one out of four) presented in recent work. Therefore sophisticated calculations for the required parameters are introduced. The provided 1oo4-architecture represents an advanced safety architecture based on on-chip redundancy, which is 3-failure safe. This means that at least one of the four channels have to work correctly in order to trigger the safety function.

  1. Calculating the cost of research and Development in nuclear and radiation safety

    International Nuclear Information System (INIS)

    Matsulevich, N.Je.; Nosovs'ka, A.A.

    2010-01-01

    Methodological support assessing the cost of research and development in the area of nuclear and radiation safety regulation is considered. Basic methodological recommendations for determining labor expenditures for research and development in nuclear and radiation safety are provided.

  2. Effects of organizational safety practices and perceived safety climate on PPE usage, engineering controls, and adverse events involving liquid antineoplastic drugs among nurses.

    Science.gov (United States)

    DeJoy, David M; Smith, Todd D; Woldu, Henok; Dyal, Mari-Amanda; Steege, Andrea L; Boiano, James M

    2017-07-01

    Antineoplastic drugs pose risks to the healthcare workers who handle them. This fact notwithstanding, adherence to safe handling guidelines remains inconsistent and often poor. This study examined the effects of pertinent organizational safety practices and perceived safety climate on the use of personal protective equipment, engineering controls, and adverse events (spill/leak or skin contact) involving liquid antineoplastic drugs. Data for this study came from the 2011 National Institute for Occupational Safety and Health (NIOSH) Health and Safety Practices Survey of Healthcare Workers which included a sample of approximately 1,800 nurses who had administered liquid antineoplastic drugs during the past seven days. Regression modeling was used to examine predictors of personal protective equipment use, engineering controls, and adverse events involving antineoplastic drugs. Approximately 14% of nurses reported experiencing an adverse event while administering antineoplastic drugs during the previous week. Usage of recommended engineering controls and personal protective equipment was quite variable. Usage of both was better in non-profit and government settings, when workers were more familiar with safe handling guidelines, and when perceived management commitment to safety was higher. Usage was poorer in the absence of specific safety handling procedures. The odds of adverse events increased with number of antineoplastic drugs treatments and when antineoplastic drugs were administered more days of the week. The odds of such events were significantly lower when the use of engineering controls and personal protective equipment was greater and when more precautionary measures were in place. Greater levels of management commitment to safety and perceived risk were also related to lower odds of adverse events. These results point to the value of implementing a comprehensive health and safety program that utilizes available hazard controls and effectively communicates

  3. Investigation of the possibility of a calculative reactor safety estimation in the licence procedure for nuclear reactors

    International Nuclear Information System (INIS)

    Adler, B.; Kampf, T.

    1975-12-01

    Up to now it is impossible to calculate completely the safety of nuclear reactors. Therefore the authors have collected and employed a number of at a high degree independent safety parameters for mathematical evaluation of the reactor safety. By means of computer programs such parameters from about 400 research reactors have been analysed and the fluctuation ranges of their greatest density were determined. The limits of these fluctuation ranges are quickly available and can be used as recommended values for the layout and for the safety estimation of research reactors. A comparison of the existing layout recommendations and the determined fluctuation ranges in most cases shows a good agreement. In some cases corrections and new layout recommendations have been proposed. The determined fluctuation ranges found their first practical application in the estimation of the Rossendorf Equipment for Critical Experiments (RAKE). (author)

  4. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  5. Patient involvement in patient safety: Protocol for developing an intervention using patient reports of organisational safety and patient incident reporting

    Directory of Open Access Journals (Sweden)

    Armitage Gerry

    2011-05-01

    Full Text Available Abstract Background Patients have the potential to provide a rich source of information on both organisational aspects of safety and patient safety incidents. This project aims to develop two patient safety interventions to promote organisational learning about safety - a patient measure of organisational safety (PMOS, and a patient incident reporting tool (PIRT - to help the NHS prevent patient safety incidents by learning more about when and why they occur. Methods To develop the PMOS 1 literature will be reviewed to identify similar measures and key contributory factors to error; 2 four patient focus groups will ascertain practicality and feasibility; 3 25 patient interviews will elicit approximately 60 items across 10 domains; 4 10 patient and clinician interviews will test acceptability and understanding. Qualitative data will be analysed using thematic content analysis. To develop the PIRT 1 individual and then combined patient and clinician focus groups will provide guidance for the development of three potential reporting tools; 2 nine wards across three hospital directorates will pilot each of the tools for three months. The best performing tool will be identified from the frequency, volume and quality of reports. The validity of both measures will be tested. 300 patients will be asked to complete the PMOS and PIRT during their stay in hospital. A sub-sample (N = 50 will complete the PMOS again one week later. Health professionals in participating wards will also be asked to complete the AHRQ safety culture questionnaire. Case notes for all patients will be reviewed. The psychometric properties of the PMOS will be assessed and a final valid and reliable version developed. Concurrent validity for the PIRT will be assessed by comparing reported incidents with those identified from case note review and the existing staff reporting scheme. In a subsequent study these tools will be used to provide information to wards/units about their

  6. The effect of safety training involving non-destructive testing among students at specialized vocational high schools

    International Nuclear Information System (INIS)

    Lim Young Khi; Han, Eun Ok; Choi, Yoon Seok

    2017-01-01

    By examining the safety issues involved in on-site training sessions conducted at specialized vocational high schools, and by analyzing the effects of non-destructive testing (NDT) safety training, this study aims to contribute to ensuring the general safety of high school students. Students who expressed an interest in participation were surveyed regarding current NDT training practices, as well as NDT safety training. A total of 361 students from 4 schools participated in this study; 37.7% (136 students) were from the Seoul metropolitan area and 62.3% (225 students) were from other areas. Of the respondents, 2.2% (8 students) reported having engaged in NDT. As a result of safety training, statistically significant improvements were observed in most areas, except for individuals with previous NDT experience. The areas of improvement included safety awareness, acquisition of knowledge, subjective knowledge levels, objective knowledge levels, and adjustments to existing personal attitudes. Even at absolutely necessary observation-only training sessions, it is crucial that sufficient safety training and additional safety measures be adequately provided

  7. The effect of safety training involving non-destructive testing among students at specialized vocational high schools

    Energy Technology Data Exchange (ETDEWEB)

    Lim Young Khi [Dept. of Radiological Science, Gachon University, Inchon (Korea, Republic of); Han, Eun Ok; Choi, Yoon Seok [Dept. of Education amd Research, Korea Academy of Nuclear Safety, Seoul (Korea, Republic of)

    2017-06-15

    By examining the safety issues involved in on-site training sessions conducted at specialized vocational high schools, and by analyzing the effects of non-destructive testing (NDT) safety training, this study aims to contribute to ensuring the general safety of high school students. Students who expressed an interest in participation were surveyed regarding current NDT training practices, as well as NDT safety training. A total of 361 students from 4 schools participated in this study; 37.7% (136 students) were from the Seoul metropolitan area and 62.3% (225 students) were from other areas. Of the respondents, 2.2% (8 students) reported having engaged in NDT. As a result of safety training, statistically significant improvements were observed in most areas, except for individuals with previous NDT experience. The areas of improvement included safety awareness, acquisition of knowledge, subjective knowledge levels, objective knowledge levels, and adjustments to existing personal attitudes. Even at absolutely necessary observation-only training sessions, it is crucial that sufficient safety training and additional safety measures be adequately provided.

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

    International Nuclear Information System (INIS)

    Maul, Philip; Robinson, Peter

    2005-03-01

    SKB have published their Interim Main Report of the safety assessment SR-Can, which is intended to establish the framework for what will be submitted in 2006 in support of a licence application for construction of the spent fuel encapsulation plant. This follows on from the SR-Can Planning Document published in 2003. The purpose of the Interim Report is stated to be to demonstrate the methodology that will be used for safety assessment. The present report evaluates the information provided in the Interim SR-Can Report that is relevant to the Performance Assessment (PA) calculations that SKB intend to undertake, using independent calculations to facilitate this process. SKB consider that the primary safety function is to isolate completely the fuel within the canisters over the entire assessment period. Should a canister be damaged, the secondary safety function is to ensure that any release is retarded and dispersed sufficiently to ensure that concentrations levels in the accessible environment cannot cause unacceptable consequences. In this report PA calculations are considered to include both a high-level representation of the evolution of the system (relevant to the primary safety function), and any subsequent radionuclide transport (relevant to the secondary safety function). The main conclusions drawn are: 1. The effects of climate evolution on engineered barriers have not been analysed in detail in the Interim Report, and this limits the usefulness of the preliminary calculations that have been undertaken. 2. A key aspect of SKB's approach is the use of an integrated near-field evolution model. The information provided on this model demonstrates its capability efficiently to reproduce calculations from individual process models, but insufficient information is given at the present time to justify statements about interactions between processes. In particular it is assumed that relatively short term thermal and resaturation processes do not affect the

  9. Radiation effect calculation means of the Crisis Technical Center of the Nuclear Safety and Protection Institut

    International Nuclear Information System (INIS)

    Crabol, B.; Manesse, D.; Robeau, D.

    1989-07-01

    The available calculation tools of the Crisis Technical Center (CTC), for the analysis and evaluation of radiation effects from a nuclear accident, are presented. The CTC calculation unit depends on local means, and on the National Meteorology system, in order to collect the data needed for the atmospheric waste diffusion evaluation. For the radiation dose calculations, plotters and software allowing the analysis of all waste Kinetics and all the meteorological conditions are available. The work developed by CTC calculation unit enables an easy application of the calculation tools as well as the results obtention. Images from data bases are provided to complete the obtained results [fr

  10. Risk-based evaluation tool for safety-related maintenance involving scaffolding

    International Nuclear Information System (INIS)

    Stevens, C.; Azizi, M.; Massman, M.

    1988-01-01

    The US Nuclear Regulatory Commission (NRC) has expressed a general concern that transient materials in and around safety systems at nuclear power plants represent a seismic safety hazard to the plant, in particular, the uncontrolled use of scaffolding during maintenance activities. Currently, most plants perform a seismic safety analysis for all uses of scaffolding near safety-related equipment to determine appropriate tie-down locations, scaffolding reinforcements, etc. This is both time-consuming and, for the most part, unnecessary. A workable engineering solution based on risk analysis techniques has been developed and is being used at the Palo Verde nuclear generating station (PVNGS)

  11. Stakeholder involvement in nuclear issues. INSAG-20. A report by the International Nuclear Safety Group

    International Nuclear Information System (INIS)

    2006-01-01

    Many of the world's nuclear power plants were constructed long ago without much public involvement in the associated decision making. It is anticipated, however, that a variety of stakeholders will seek participation in such decisions now as the nuclear option is being revisited in many places. Accidents at Three Mile Island and Chernobyl, among other places, have served to arouse public concern. The development of 'here-and-now' media capabilities has created an awareness that may not have previously existed. Improvements in educational systems and the development of the Internet have made technical information and expertise available to individuals and locations that were previously without them. In addition, consideration of the environmental impacts of various energy strategies has moved to the fore. INSAG has concluded that the expectations of stakeholders of a right to participate in energy decisions are something that the nuclear community must address. Decisions regarding such matters as the siting and construction of a nuclear power plant are no longer largely the domain of a closed community of technical experts and utility executives. Today, the concerns and expectations of all manner of persons and organizations - from the local farmer to the international financial institution - must be considered. This report is intended for use by all stakeholders in the nuclear community - national regulatory authorities, nuclear power plant designers and operators, public interest organizations and individuals, the media and, not to be forgotten, local and national populations. INSAG's fundamental conclusion is that all stakeholders with an interest in nuclear decisions should be provided with an opportunity for full and effective participation in them. With this right, however, come certain obligations on all sides for openness, candour and civility. INSAG is hopeful that this report will help define the interests and roles of the stakeholders in the nuclear

  12. [The experience of involvement of volunteers into maintenance of infection safety during period of implementation of mass activities].

    Science.gov (United States)

    Imamov, A A; Balabanova, L A; Zamalieva, M A

    2016-01-01

    The article presents experience of Rospotrebnadzor in the Republic of Tatarstan in the field of preventive medicine concerning training of volunteers on issues of infection safety with purpose of prevention of ictuses of infection diseases during mass activities with international participation in the period of XXVII World Summer Students Games. The model of hygienic training for volunteers provides two directions: training for volunteers ’ leaders on issues of infection safety and remote course for involved volunteers. During period of preparation for the Students Games-2013 hygienic training was organized for volunteers-leaders in the field of infection safety with following attestation. The modern training technologies were applied. The volunteers-leaders familiarized with groups of infection diseases including the most dangerous ones, investigated with expert algorithm of actions to be applied in case of suspicion on infection disease in gest or participant of the Games-2013 to secure one's health and health of immediate population. The active volunteers-leaders became trainers and coaches in the field of infection safety. The second stage of infection safety training organized by youth trainers' pool in number of 30 individuals the training technology "Equal trains equal" was applied for hygienic training of volunteers involved at epidemiologically significant objects (food objects, hotels, accompaniment of guests and sportsmen). The volunteers-leaders trained to infection safety 1400 volunteers. The format of electronic personal cabinet and remote course were selected as tools of post-training monitoring.

  13. Patient safety: numerical skills and drug calculation abilities of nursing students and registered nurses.

    Science.gov (United States)

    McMullan, Miriam; Jones, Ray; Lea, Susan

    2010-04-01

    This paper is a report of a correlational study of the relations of age, status, experience and drug calculation ability to numerical ability of nursing students and Registered Nurses. Competent numerical and drug calculation skills are essential for nurses as mistakes can put patients' lives at risk. A cross-sectional study was carried out in 2006 in one United Kingdom university. Validated numerical and drug calculation tests were given to 229 second year nursing students and 44 Registered Nurses attending a non-medical prescribing programme. The numeracy test was failed by 55% of students and 45% of Registered Nurses, while 92% of students and 89% of nurses failed the drug calculation test. Independent of status or experience, older participants (> or = 35 years) were statistically significantly more able to perform numerical calculations. There was no statistically significant difference between nursing students and Registered Nurses in their overall drug calculation ability, but nurses were statistically significantly more able than students to perform basic numerical calculations and calculations for solids, oral liquids and injections. Both nursing students and Registered Nurses were statistically significantly more able to perform calculations for solids, liquid oral and injections than calculations for drug percentages, drip and infusion rates. To prevent deskilling, Registered Nurses should continue to practise and refresh all the different types of drug calculations as often as possible with regular (self)-testing of their ability. Time should be set aside in curricula for nursing students to learn how to perform basic numerical and drug calculations. This learning should be reinforced through regular practice and assessment.

  14. Results from synthesis of calculation cases illustrating overall system performance in the safety assessment in H12 report

    International Nuclear Information System (INIS)

    Makino, Hitoshi; Sawada, Atsushi; Wakasugi, Keiichiro; Kato, Tomoko; Uchida, Masahiro; Miyahara, Kaname

    2002-02-01

    JNC (Japan Nuclear Cycle Development Institute) had proceeded R and D activities to provide a scientific and technical basis for geological disposal of HLW in Japan. The second progress report (H12) documented the progress of R and D and the Japanese version was submitted to the AEC (the Atomic Energy Commission) in November 1999. This report summarizes the calculation results for nuclide migration in 'Synthesis of Calculation Cases Illustrating Overall System Performance', which are performed to examine the safety of the geological disposal concept in Japan in the Safety Assessment in H12 Report. In addition, a set of calculation result for nuclide migration through each pathway in one-dimensional multiple pathway model (a set of 48 segments) are summarized for the Reference Case in H12 Report, and calculated dose conversion factors are also summarized against the combinations of potential Geosphere-Biosphere Interfaces (GBI) and potential exposure groups. Digital data of the calculation results are summarized in Appendix CD-ROM as Microsoft EXCEL files. (author)

  15. 49 CFR 244.13 - Subjects to be addressed in a Safety Integration Plan involving an amalgamation of operations.

    Science.gov (United States)

    2010-10-01

    ... transaction: (a) Corporate culture. Each applicant shall: (1) Identify and describe differences for each safety-related area between the corporate cultures of the railroads involved in the transaction; (2... step-by-step measures, the integration of these corporate cultures and the manner in which it will...

  16. Standardized dose factors for dose calculations - 1982 SRP reactor safety analysis report tritium, iodine, and noble gases

    International Nuclear Information System (INIS)

    Pillinger, W.L.; Marter, W.L.

    1982-01-01

    Standardized dose constants are recommended for calculation of offsite doses in the 1982 SRP Reactor Safety Analysis Report (SAR). Dose constants are proposed for inhalation of tritium and radioiodines and for submersion in a semi-infinite cloud of radioiodines and noble gases. The proposed constants, based on ICRP2 methodology for internal dose and methodology recommended by the US Nuclear Regulatory Commission for external dose, are compatible with dose calculational methods used at the Savannah River Plant and Savannah River Laboratory for normal releases of radioactivity. 8 references

  17. Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2010-12-01

    Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

  18. Nursing involvement in risk and patient safety management in Primary Care.

    Science.gov (United States)

    Coronado-Vázquez, Valle; García-López, Ana; López-Sauras, Susana; Turón Alcaine, José María

    Patient safety and quality of care in a highly complex healthcare system depends not only on the actions of professionals at an individual level, but also on interaction with the environment. Proactive risk management in the system to prevent incidents and activities targeting healthcare teams is crucial in establishing a culture of safety in centres. Nurses commonly lead these safety strategies. Even though safety incidents are relatively infrequent in primary care, since the majority are preventable, actions at this level of care are highly effective. Certification of services according to ISO standard 9001:2008 focuses on risk management in the system and its use in certifying healthcare centres is helping to build a safety culture amongst professionals. Copyright © 2017 Elsevier España, S.L.U. All rights reserved.

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

    International Nuclear Information System (INIS)

    Ivanov, Evgeny; Ivanova, Tatiana; Pignet, Sophie

    2013-01-01

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

  20. Annual progress report FY 1977. [Computer calculations of light water reactor dynamics and safety

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, K.F.; Henry, A.F.

    1977-07-01

    Progress is summarized in a project directed toward development of numerical methods suitable for the computer solution of problems in reactor dynamics and safety. Specific areas of research include methods of integration of the time-dependent diffusion equations by finite difference and finite element methods; representation of reactor properties by various homogenization procedures; application of synthesis methods; and development of response matrix techniques.

  1. Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

    International Nuclear Information System (INIS)

    Blideanu, Valentin; Garcia, Mauricio; Joyer, Philippe; Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco; Ortiz, Felix; Sanz, Javier; Sauvan, Patrick

    2011-01-01

    In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

  2. Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

    Energy Technology Data Exchange (ETDEWEB)

    Blideanu, Valentin [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Garcia, Mauricio [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Joyer, Philippe, E-mail: philippe.joyer@cea.fr [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Ortiz, Felix [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Sanz, Javier; Sauvan, Patrick [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain)

    2011-10-01

    In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

  3. Hanford 300 Area Treated Effluent Disposal Facility inventory at risk calculations and safety analysis

    International Nuclear Information System (INIS)

    Olander, A.R.

    1995-11-01

    The 300 Area Treated Effluent Disposal Facility (TEDF) is a wastewater treatment plant being constructed to treat the 300 Area Process Sewer and Retention Process Sewer. This document analyzes the TEDF for safety consequences. It includes radionuclide and hazardous chemical inventories, compares these inventories to appropriate regulatory limits, documents the compliance status with respect to these limits, and identifies administrative controls necessary to maintain this status

  4. Mathematical calculation skills required for drug administration in undergraduate nursing students to ensure patient safety: A descriptive study: Drug calculation skills in nursing students.

    Science.gov (United States)

    Bagnasco, Annamaria; Galaverna, Lucia; Aleo, Giuseppe; Grugnetti, Anna Maria; Rosa, Francesca; Sasso, Loredana

    2016-01-01

    In the literature we found many studies that confirmed our concerns about nursing students' poor maths skills that directly impact on their ability to correctly calculate drug dosages with very serious consequences for patient safety. The aim of our study was to explore where students had most difficulty and identify appropriate educational interventions to bridge their mathematical knowledge gaps. This was a quali-quantitative descriptive study that included a sample of 726 undergraduate nursing students. We identified exactly where students had most difficulty and identified appropriate educational interventions to bridge their mathematical knowledge gaps. We found that the undergraduate nursing students mainly had difficulty with basic maths principles. Specific learning interventions are needed to improve their basic maths skills and their dosage calculation skills. For this purpose, we identified safeMedicate and eDose (Authentic World Ltd.), only that they are only available in English. In the near future we hope to set up a partnership to work together on the Italian version of these tools. Copyright © 2015 Elsevier Ltd. All rights reserved.

  5. Specification of materials Data for Fire Safety Calculations based on ENV 1992-1-2

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl

    1997-01-01

    of constructions of any concrete exposed to any time of any fire exposure can be calculated.Chapter 4.4 provides information on what should be observed if more general calculation methods are used.Annex A provides some additional information on materials data. This chapter is not a part of the code......The part 1-2 of the Eurocode on Concrete deals with Structural Fire Design.In chapter 3, which is partly written by the author of this paper, some data are given for the development of a few material parameters at high temperatures. These data are intended to represent the worst possible concrete...... to experience form tests on structural specimens based on German siliceous concrete subjected to Standard fire exposure until the time of maximum gas temperature.Chapter 4.3, which is written by the author of this paper, provides a simplified calculation method by means of which the load bearing capacity...

  6. Heavy-Particle Collisions Involving Many Active Electrons: How (In-)Accurate Are Our Calculated Cross Sections?

    International Nuclear Information System (INIS)

    Kirchner, Tom

    2014-01-01

    Full text: The theoretical description of ion-atom and ion-molecule collisions is a difficult task: one deals with a two-center or a multi-center problem, for which standard angular momentum expansions do not work very well, and one typically faces the problem that several processes, such as electron transfer and ionization into the continuum, compete with each other. If more than two electrons are present, the numerical solution of the full Schrödinger equation of the collision system is out of reach and assumptions and approximations have to be introduced at the outset. This is to say that one solves (at most) a model in order to describe the collision system and, as a consequence, has to deal with a two-fold problem when it comes to estimating the uncertainties and inaccuracies of the calculated data: (i) to assess the limitations of the model (which may be compared with quantifying systematic errors in an experiment); (ii) to perform careful convergence studies for the numerical procedures involved (which may be compared with narrowing statistical experimental errors). These two interrelated problems were illustrated by using a recent work on X-ray emission from a highly-charged ion after electron capture as an example. The calculations for this problem are based on the assumption that collisional capture and post-collisional de-excitation processes can be treated independently. This introduces a first systematic error, but probably a very small one, because capture and de-excitation take place on different time scales. Similarly, the assumption of a classical straight-line projectile trajectory is uncritical. Three sources of significant uncertainties are present in the collision calculation: (i) usage of the independent-electron model, (ii) usage of a finite basis set to solve the single-electron time-dependent Schrödinger equation, (iii) usage of multinomial statistics to calculate multiple (shell-specific) capture probabilities, which form the starting

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  8. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements

    International Nuclear Information System (INIS)

    Domingos, Douglas Borges

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a Miniplate Irradiation Device (MID) to be placed in the IEA-R1 reactor core. The irradiation device is used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 - Al dispersion fuels, LEU type (19.75 % 235 U) with uranium densities of, respectively, 3.2 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and 2DB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  9. The Influence Paths of Emotion on the Occupational Safety of Rescuers Involved in Environmental Emergencies- Systematic Review Article.

    Science.gov (United States)

    Lu, Jintao; Yang, Naiding; Ye, Jinfu; Wu, Haoran

    2014-11-01

    A detailed study and analysis of previous research has been carried out to illustrate the relationships between a range of environmental emergencies, and their effects on the emotional state of the rescuers involved in responding to them, by employing Pub Med, Science Direct, Web of Science, Google Scholar, CNKI and Scopus for required information with the several keywords "emergency rescue", "occupational safety", "natural disaster", "emotional management". The effect of the rescuers' emotion on their occupational safety and immediate and long-term emotional behavior is then considered. From these considerations, we suggested four research propositions related to the emotional effects at both individual and group levels, and to the responsibilities of emergency response agencies in respect of ensuring the psychological and physical occupational safety of rescuers during and after environmental emergencies. An analysis framework is proposed which could be used to study the influence paths of these different aspects of emotional impact on a range of occupational safety issues for rescue workers. The authors believe that the conclusions drawn in this paper can provide a useful theoretical reference for decision-making related to the management and protection of the occupational safety of rescuers responding to natural disasters and environmental emergencies.

  10. X/Qs and unit dose calculations for Central Waste Complex interim safety basis effort

    International Nuclear Information System (INIS)

    Huang, C.H.

    1996-01-01

    The objective for this problem is to calculate the ground-level release dispersion factors (X/Q) and unit doses for onsite facility and offsite receptors at the site boundary and at Highway 240 for plume meander, building wake effect, plume rise, and the combined effect. The release location is at Central Waste Complex Building P4 in the 200 West Area. The onsite facility is located at Building P7. Acute ground level release 99.5 percentile dispersion factors (X/Q) were generated using the GXQ. The unit doses were calculated using the GENII code. The dimensions of Building P4 are 15 m in W x 24 m in L x 6 m in H

  11. Calculational framework for safety analyses of non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Coleman, J.R.

    1994-01-01

    A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks

  12. Radioactive waste storage facilities, involvement of AVN in inspection and safety assessment

    International Nuclear Information System (INIS)

    Simenon, R.; Smidts, O.

    2006-01-01

    The legislative and regulatory framework in Belgium for the licensing and the operation of radioactive waste storage buildings are defined by the Royal Decree of 20 July 2001 (hereby providing the general regulations regarding to the protection of the population, the workers and the environment against the dangers of ionising radiation). This RD introduces in the Belgian law the radiological protection and ALARA-policy concepts. The licence of each nuclear facility takes the form of a Royal Decree of Authorization. It stipulates that the plant has to be in conformity with its Safety Analysis Report. This report is however not a public document but is legally binding. Up to now, the safety assessment for radioactive waste storage facilities, which is implemented in this Safety Analysis Report, has been judged on a case-by-case basis. AVN is an authorized inspection organisation to carry out the surveillance of the Belgian nuclear installations and performs hereby nuclear safety assessments. AVN has a role in the nuclear safety and radiation protection during all the phases of a nuclear facility: issuance of licenses, during design and construction phase, operation (including reviewing and formal approval of modifications) and finally the decommissioning. Permanent inspections are performed on a regular basis by AVN, this by a dedicated site inspector, who is responsible for a site of an operator with nuclear facilities. Besides the day-to-day inspections during operation there are also the periodic safety reviews. AVN assesses the methodological approaches for the analyses, reviews and approves the final studies and results. The conditioned waste in Belgium is stored on the Belgoprocess' sites (region Mol-Dessel) for an intermediate period (about 80 years). In the meantime, a well-defined inspection programme is being implemented to ensure that the conditioned waste continues to be stored safely during this temporary storage period. This programme was draw up by

  13. Landscape modeling for dose calculations in the safety assessment of a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Lindborg, Tobias; Kautsky, Ulrik; Brydsten, Lars

    2007-01-01

    The Swedish Nuclear Fuel and Waste Management Co.,(SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site. Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the

  14. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J M; Kodeli, I; Menard, St; Bouchet, J L; Renard, F; Martin, E; Blazy, L; Voros, S; Bochud, F; Laedermann, J P; Beaugelin, K; Makovicka, L; Quiot, A; Vermeersch, F; Roche, H; Perrin, M C; Laye, F; Bardies, M; Struelens, L; Vanhavere, F; Gschwind, R; Fernandez, F; Quesne, B; Fritsch, P; Lamart, St; Crovisier, Ph; Leservot, A; Antoni, R; Huet, Ch; Thiam, Ch; Donadille, L; Monfort, M; Diop, Ch; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

  15. Stakeholder involvement in building and maintaining radiation safety infrastructure in Latvia: The case studies

    International Nuclear Information System (INIS)

    Eglajs, A.; Salmins, A.

    2003-01-01

    This paper comprises the assessment of interests for central and local governments, different authorities, public and commercial companies, political parties and non-governmental organizations, organised and ad-hock groups of public, which could contribute to development and maintenance of infrastructure for radiation safety, general environmental protection, as well as for public health among other similar fields. Understanding of these interests allows to be prepared for eventual demonstrations or publications against decisions about significant modifications of infrastructure and provides ideas how to explain needs of financial and human resources for maintaining of supervisory system and management of major facilities, which are vital for safety infrastructure. Two case studies are presented in this report related to modification of the framework law and the preparation of radioactive waste management strategy. (author)

  16. Game Theoretic Analysis of Road User Safety Scenarios Involving Autonomous Vehicles

    OpenAIRE

    Michieli, Umberto; Badia, Leonardo

    2018-01-01

    Interactions between pedestrians, bikers, and human-driven vehicles have been a major concern in traffic safety over the years. The upcoming age of autonomous vehicles will further raise major problems on whether self-driving cars can accurately avoid accidents; on the other hand, usability issues arise on whether human-driven cars and pedestrian can dominate the road at the expense of the autonomous vehicles which will be programmed to avoid accidents. This paper proposes some game theoretic...

  17. An example demonstrating the conservatism of pipe calculations using KTA safety standard 3201.2

    International Nuclear Information System (INIS)

    Zeitner, W.

    1991-01-01

    The conservatism of the code calculation is demonstrated by using an example of a highly stressed pipe subject to internal pressure and a dynamic bending moment. For this reason the allowable code loadings are compared with the load carrying capacity, which is derived by realistic analysis (plastic strains) and experiment. The latter analysis is based on measured stress-strain curves of materials and stresses at which crack initiation occurs. The experiment shows that the pipe is capable of withstanding considerably higher loads than the code permits. The realistic analysis explains this discrepancy. (orig.)

  18. Calculation of fluid-structure interaction for reactor safety with the Cassiopee code

    International Nuclear Information System (INIS)

    Graveleau, J.L.; Louvet, P.D.

    1979-01-01

    The cassiopee code is an eulerian-lagrangian coupled code for computations where the hydrodynamic is coupled with structural domains. It is completely explicit. The fluid zones may be computed either in lagrangian or in eulerian coordinates; thin shells can be computed wih their flexural behaviour; elastic plastic zones must be calculated in a lagrangian way. This code is under development in Cadarache. Its purpose is to compute the hypothetical core disruptive accident of a LMFBR when lagrangian codes are not sufficient. This paper contains a description of the code and two examples of computations, one of which has been compared with experimental results

  19. The medical student as a patient: attitudes towards involvement in the quality and safety of health care.

    Science.gov (United States)

    Davis, Rachel E; Joshi, Devavrata; Patel, Krishan; Briggs, M; Vincent, Charles A

    2013-10-01

    In recent years, factors that affect patients' willingness and ability to participate in safety-relevant behaviours have been investigated. However, how trained healthcare professionals or medical students would feel participating in safety-relevant behaviours as a patient in hospital remains largely unexplored. To investigate medical students' willingness to participate in behaviours related to the quality and safety of their health care. A cross-sectional exploratory study using a survey that addressed willingness to participate in different behaviours recommended by current patient safety initiatives. Three types of interactional behaviours (asking factual or challenging questions, notifying doctors or nurses of errors/problems) and three non-interactional behaviours (choosing a hospital based on the safety record, bringing medicines and a list of allergies into hospital, and reporting an error to a national reporting system) were assessed. One hundred and seventy-nine medical students from an inner city London teaching hospital participated in the study. Students' willingness to participate was affected (P interactional behaviours) whether the patient was engaging in the specific action with a doctor or nurse. Students were least willing to ask 'challenging' questions to doctors and nurses and to report errors to a national reporting system. Doctors' and nurses' encouragement appeared to increase self-reported willingness to participate in behaviours where baseline willingness was low. Similar to research on lay patient populations; medical students do not view involvement in safety-related behaviours equally. Interventions should be tailored at encouraging students to participate in behaviours they are less inclined to take on an active role in. Future research is required to examine students' motivations for participation in this important but heavily under-researched area. © 2012 John Wiley & Sons Ltd.

  20. Workers' involvement--a missing component in the implementation of occupational safety and health management systems in enterprises.

    Science.gov (United States)

    Podgórski, Daniel

    2005-01-01

    Effective implementation of occupational safety and health (OSH) legislation based on European Union directives requires promotion of OSH management systems (OSH MS). To this end, voluntary Polish standards (PN-N-18000) have been adopted, setting forth OSH MS specifications and guidelines. However, the number of enterprises implementing OSH MS has increased slowly, falling short of expectations, which call for a new national policy on OSH MS promotion. To develop a national policy in this area, a survey was conducted in 40 enterprises with OSH MS in place. The survey was aimed at identifying motivational factors underlying OSH MS implementation decisions. Specifically, workers' and their representatives' involvement in OSH MS implementation was investigated. The results showed that the level of workers' involvement was relatively low, which may result in a low effectiveness of those systems. The same result also applies to the involvement of workers' representatives and that of trade unions.

  1. Organization and liability of British regulating authorities involved in nuclear safety and radiation protection

    International Nuclear Information System (INIS)

    Harbison, S.

    1995-01-01

    In Great Britain, nuclear safety juridic basis is made of two law: HSWA (1974) for hygiene and security in working environment, and NIA (1965) specific to nuclear sites. The HSWA law created an HSC (Hygiene and Security Commission) in charge of workers and public security. HSC executive organ is HSE, whose nuclear office is NSD. Nevertheless, the general philosophy remains the one of HSWA, which results in the liability of operators in nuclear matters, as well as for any other industrial matter. (D.L.). 1 fig., 1 map

  2. Experimental validation of calculated capture rate for nucleus involved in fuel cycle; Validation experimentale du calcul du taux de capture des noyaux intervenant dans le cycle du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Benslimane-Bouland, A

    1997-09-01

    The framework of this study was the evaluation of the nuclear data requirements for Actinides and Fission Products applied to current nuclear reactors as well as future applications. This last item includes extended irradiation campaigns, 100 % Mixed Oxide fuel, transmutation or even incineration. The first part of this study presents different types of integral measurements which are available for capture rate measurements, as well as the methods used for reactor core calculation route design and nuclear data library validation. The second section concerns the analysis of three specific irradiation experiments. The results have shown the extent of the current knowledge on nuclear data as well as the associated uncertainties. The third and last section shows both the coherency between all the results, and the statistical method applied for nuclear data library adjustment. A relevant application of this method has demonstrated that only specifically chosen integral experiments can be of use for the validation of nuclear data libraries. The conclusion is reached that even if co-ordinated efforts between reactor and nuclear physicists have made possible a huge improvement in the knowledge of capture cross sections of the main nuclei such as uranium and plutonium, some improvements are currently necessary for the minor actinides (Np, Am and Cm). Both integral and differential measurements are recommended to improve the knowledge of minor actinide cross sections. As far as integral experiments are concerned, a set of criteria to be followed during the experimental conception have been defined in order to both reduce the number of required calculation approximations, and to increase as much as possible the maximum amount of extracted information. (author)

  3. Balancing Autonomy Rights and Protection: Children's Involvement in a Child Safety Online Project

    Science.gov (United States)

    Ost, Suzanne

    2013-01-01

    Researchers who involve children in their research are faced with the challenge of choosing between differing theoretical approaches which can prioritise children's autonomy rights or their "vulnerability" and their need to be protected. Somewhat confusingly, ethical guidelines seem to reflect a combination of these approaches. Even when…

  4. Criticality safety studies involved in actions to improve conditions for storing 'RA' research reactor spent fuel

    International Nuclear Information System (INIS)

    Matausek, M.; Marinkovic, N.

    1998-01-01

    A project has recently been initiated by the VINCA Institute of Nuclear Sciences to improve conditions in the spent fuel storage pool at the 6.5 MW research reactor RA, as well as to consider transferring this spent fuel into a new dry storage facility built for the purpose. Since quantity and contents of fissile material in the spent fuel storage at the RA reactor are such that possibility of criticality accident can not be a priori excluded, according to standards and regulations for handling fissile material outside a reactor, before any action is undertaken subcriticality should be proven under normal, as well as under credible abnormal conditions. To perform this task, comprehensive nuclear criticality safety studies had to be performed. (author)

  5. Modelling of contact problems involved in ensuring the safety of rail transport

    Directory of Open Access Journals (Sweden)

    Edward Rydygier

    2013-12-01

    Full Text Available Background: Mathematical modelling aids diagnostics the track and rolling stock, as it often for technical reasons it is not possible to obtain a complete set of measurement data required to diagnose the rail and wheel deformation caused by the impact of a rail vehicle on the track. The important issue in a railway diagnostics is to study the effects of contact wheel and rail. Diagnostics investigations of track and rolling stock have a fundamental role in ensuring the safety of transport of passengers and goods. The aim of the study presented in the paper was to develop simulation methods of mathematical modelling of the wheel-rail system useful in the diagnostics of the track and a railway vehicle. Methods: In the paper two ways of modelling were presented and discussed. One of these ways is the method which consists in reducing the contact issue to field issue and solving the identification of the field source in 2-D system. Also presented a different method designed on the basis of the methods using one period energy concept. This method is adapted for modelling the dynamics of the contact wheel-rail for the normal force. It has been shown that the developed modelling methods to effectively support the study on the effects of mechanical and thermal of contact wheel-rail and contribute to the safety of operations.  Results and conclusions:  In the case of field sources identifications two specific issues were examined: the issue of rail torsion and the identification of heat sources in the rail due to exposure the rolling contact wheel-rail. In the case of the method using one period energy concept it was demonstrated the usefulness of this method to the study of energy processes in the contact wheel-rail under the normal periodic force. The future direction of research is to establish cooperation with research teams entrusted with the diagnostic measurements of track and rolling stock.  

  6. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  7. Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

    International Nuclear Information System (INIS)

    Vlasichev, G.N.

    1994-01-01

    Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs

  8. Comparison of model Hartree-Fock type calculation schemes involving various non-degenerate and quasi-degenerate intrinsic Hamiltonians

    International Nuclear Information System (INIS)

    Amusa, A.

    1983-03-01

    Different Hamiltonians and their corresponding rotationally degenerate intrinsic counterparts are employed in the study of 18 O nucleus under the normal Hartree-Fock, as well as under six other Hartree-Fock type variational calculation schemes. The results are compared and then assessed in the light of their closeness or otherwise to the full 1s-0d basis shell model calculations for this nucleus. The use of these schemes for other shells is also considered. (author)

  9. Soldering and brazing safety guide: A handbook on space practice for those involved in soldering and brazing

    Science.gov (United States)

    This manual provides those involved in welding and brazing with effective safety procedures for use in performance of their jobs. Hazards exist in four types of general soldering and brazing processes: (1) cleaning; (2) application of flux; (3) application of heat and filler metal; and (4) residue cleaning. Most hazards during those operations can be avoided by using care, proper ventilation, protective clothing and equipment. Specific process hazards for various methods of brazing and soldering are treated. Methods to check ventilation are presented as well as a check of personal hygiene and good maintenance practices are stressed. Several emergency first aid treatments are described.

  10. Improvement of Modeling Scheme of the Safety Injection Tank with Fluidic Device for Realistic LBLOCA Calculation

    International Nuclear Information System (INIS)

    Bang, Young Seok; Cheong, Aeju; Woo, Sweng Woong

    2014-01-01

    Confirmation of the performance of the SIT with FD should be based on thermal-hydraulic analysis of LBLOCA and an adequate and physical model simulating the SIT/FD should be used in the LBLOCA calculation. To develop such a physical model on SIT/FD, simulation of the major phenomena including flow distribution of by standpipe and FD should be justified by full scale experiment and/or plant preoperational testing. Author's previous study indicated that an approximation of SIT/FD phenomena could be obtained by a typical system transient code, MARS-KS, and using 'accumulator' component model, however, that additional improvement on modeling scheme of the FD and standpipe flow paths was needed for a reasonable prediction. One problem was a depressurizing behavior after switchover to low flow injection phase. Also a potential to release of nitrogen gas from the SIT to the downstream pipe and then reactor core through flow paths of FD and standpipe has been concerned. The intrusion of noncondensible gas may have an effect on LBLOCA thermal response. Therefore, a more reliable model on SIT/FD has been requested to get a more accurate prediction and a confidence of the evaluation of LBLOCA. The present paper is to discuss an improvement of modeling scheme from the previous study. Compared to the existing modeling, effect of the present modeling scheme on LBLOCA cladding thermal response is discussed. The present study discussed the modeling scheme of SIT with FD for a realistic simulation of LBLOCA of APR1400. Currently, the SIT blowdown test can be best simulated by the modeling scheme using 'pipe' component with dynamic area reduction. The LBLOCA analysis adopting the modeling scheme showed the PCT increase of 23K when compared to the case of 'accumulator' component model, which was due to the flow rate decrease at transition phase low flow injection and intrusion of nitrogen gas to the core. Accordingly, the effect of SIT/FD modeling

  11. Trusting telemedicine: A discussion on risks, safety, legal implications and liability of involved stakeholders.

    Science.gov (United States)

    Parimbelli, E; Bottalico, B; Losiouk, E; Tomasi, M; Santosuosso, A; Lanzola, G; Quaglini, S; Bellazzi, R

    2018-04-01

    The main purpose of the article is to raise awareness among all the involved stakeholders about the risks and legal implications connected to the development and use of modern telemedicine systems. Particular focus is given to the class of "active" telemedicine systems, that imply a real-world, non-mediated, interaction with the final user. A secondary objective is to give an overview of the European legal framework that applies to these systems, in the effort to avoid defensive medicine practices and fears, which might be a barrier to their broader adoption. We leverage on the experience gained during two international telemedicine projects, namely MobiGuide (pilot studies conducted in Spain and Italy) and AP@home (clinical trials enrolled patients in Italy, France, the Netherlands, United Kingdom, Austria and Germany), whose development our group has significantly contributed to in the last 4 years, to create a map of the potential criticalities of active telemedicine systems and comment upon the legal framework that applies to them. Two workshops have been organized in December 2015 and March 2016 where the topic has been discussed in round tables with system developers, researchers, physicians, nurses, legal experts, healthcare economists and administrators. We identified 8 features that generate relevant risks from our example use cases. These features generalize to a broad set of telemedicine applications, and suggest insights on possible risk mitigation strategies. We also discuss the relevant European legal framework that regulate this class of systems, providing pointers to specific norms and highlighting possible liability profiles for involved stakeholders. Patients are more and more willing to adopt telemedicine systems to improve home care and day-by-day self-management. An essential step towards a broader adoption of these systems consists in increasing their compliance with existing regulations and better defining responsibilities for all the

  12. FLIGHT SAFETY CONTROL OF THE BASIS OF UNCERTAIN RISK EVALUATION WITH NON-ROUTINE FLIGHT CONDITIONS INVOLVED

    Directory of Open Access Journals (Sweden)

    2016-01-01

    Full Text Available The article deals with methods of forecasting the level of aviation safety operation of aircraft systems on the basis of methods of evaluation the risks of negative situations as a consequence of a functional loss of initial properties of the system with critical violations of standard modes of the aircraft. Mathematical Models of Risks as a Danger Measure of Discrete Random Events in Aviation Systems are presented. Technological Schemes and Structure of Risk Control Proce- dures without the Probability are illustrated as Methods of Risk Management System in Civil Aviation. The assessment of the level of safety and quality and management of aircraft, made not only from the standpoint of reliability (quality and consumer properties, but also from the position of ICAO on the basis of a risk-based approach. According to ICAO, the security assessment is performed by comparing the calculated risk with an acceptable level. The approach justifies the use of qualitative evaluation techniques safety in the forms of proactive forecasted and predictive risk management adverse impacts to aviation operations of various kinds, including the space sector and nuclear energy. However, for the events such as accidents and disasters, accidents with the aircraft, fighters in a training flight, during the preparation of the pilots on the training aircraft, etc. there is no required statistics. Density of probability distribution (p. d. f. of these events are only hypothetical, unknown with "hard tails" that completely eliminates the application of methods of confidence intervals in the traditional approaches to the assessment of safety in the form of the probability analysis.

  13. Stakeholder involvement in building and maintaining national and international radiation safety infrastructures

    International Nuclear Information System (INIS)

    Shimomura, K.

    2004-01-01

    Society's expectations with regard to policy towards risky technologies have changed significantly over the past 50 years, and perhaps most dramatically, over the past decade. Arrangements for the development and implementation of such policy may well fit with traditional theories from the disciplines of law, political science and engineering regarding democratic legitimacy, the delegation of power and the role of the expert. They may, however, no longer fit with a policy environment that is considerably more complex than those theories allow. The stakes are high for the radiation protection community as it seeks to recognize and accommodate these changed and changing expectations.For many years, the OECD Nuclear Energy Agency and its Committee on Radiation Protection and Public Health (CRPPH) has an active work programme on details and implications of stakeholder involvement in radiological protection decision making processes. The series of workshops in Villigen, Switzerland (in 1998 and 2002) and related follow-up work, offer assistance to the international radiological protection community on how to better integrate radiological protection into modern society. The lessons that have been learned in this area carry implications on national policy and on the governmental infrastructures necessary to carry it out

  14. About the use of the CATHARE code for best estimate Large Break LOCA calculations: benefits for Safety and constraints

    International Nuclear Information System (INIS)

    Vacher, J.L.

    1994-01-01

    Since 1979, EDF has participated to the development of CATHARE, a best estimate accidental thermalhydraulic code, in collaboration with CEA and FRAMATOME. EDF is now investigating the use of this code for licensing studies and particularly for Large Break LOCA calculations. Until now, the work done at EDF, in relation with FRAMATOME and CEA, has mainly focused on the physical analysis of the transient and on the identification of the key phenomena. This task is a necessary step before uncertainty evaluation. To illustrate this point of view, a peculiar example of calculated Large Break transients for a 900 MW three loop plant is presented. In one of these calculations, a high value of Peak Cladding Temperature was obtained. This peculiar scenario was initiated by a large entrainment of water to the steam generators at the very beginning of the reflooding stage, followed by a strong pressurization which led to a lasting draining of the reactor vessel. The physical phenomena which determine the existence and amplitude of this scenario were identified and their influence was explained: condensation at the accumulator injection, heat exchange in the core, entrainment process to the steam generators. It appeared obvious that the large observed uncertainty was associated to only a few parameters. Although this peculiar system behaviour was obtained for only a particular combination of parameters and a narrow range of thermalhydraulic conditions, the capability of the code to simulate these phenomena was investigated in regard to experimental data. It was concluded that this scenario was definitely unrealistic on a reactor. Nevertheless, this peculiar example tends to demonstrate, firstly, that the use of a best-estimate code improves Safety as it makes possible to point out physical phenomena that could not be considered when using non mechanistic codes, secondly, that the uncertainty evaluation must be guided by a pertinent physical analysis of the transient, focusing

  15. Maximising workforce involvement in HSE (Health, Safety and Environment) case development

    Energy Technology Data Exchange (ETDEWEB)

    Vine, Mark D. (DNV Energy Middle East); Ingvarson, Johan [DNV Energy, Oslo (Norway)

    2008-07-01

    The development of HSE Cases demonstrating that HSE and particularly major accident hazard risks are being managed to an acceptable level is a requirement posed on many Oil and Gas facilities by regulators and stakeholders. In order to develop a demonstration case rather than a description and to ensure a truly live document, GASCO, one of the world's largest gas processing companies, turned to DNV to develop HSE Cases for all their facilities. The DNV approach included the use of Hazard and Effects Management Process (HEMP) and the development of a customised intranet solution to ensure a continuously living process. HEMP is a systematic method of identifying hazards, assessing risk, putting controls in place to guard against those risks and defining recovery measures should an incident happen. With GASCO, HEMP was implemented through workshops involving operation workforce from senior managers to supervisory level. A powerful and popular output of HEMP is the Bow Tie diagram which is used to graphically display HSE Critical Elements and the HSE Critical Activities and Tasks (i.e. operational systems) that support them. To ensure a living process, a dedicated GASCO intranet solution was developed where all users can easily access and contribute to the electronic HSE Case without any geographical constraints. The electronic case widens the user base and creates awareness among the workforce. The added value of the HEMP and intranet solution approach is that it focuses on capitalizing on knowledge and expertise present with those operating the facilities using a 'Cradle to the Grave' approach. It is essential to maximize the interface with the workforce in order to develop and maintain a comprehensive HSE Case. (author)

  16. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

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

  17. Simplified calculation of investment costs involved in purifying industrial waste water. Calculo simplificado de los costes de inversion en la depuracion de aguas residuales industriales

    Energy Technology Data Exchange (ETDEWEB)

    Queralt, R. (Junta de Saneamientos. Generalidad de Catalua (Spain))

    1993-03-01

    The calculation of the investment involved in purifying industrial waste water poses certain problems since this is affected either by employing complicated methods which require a great deal of data or, as the sole alternative, through subjective estimates. The present article purposes an intermediate system based on simplified formulas for which it is only necessary to know three parameters, namely, (in the majority of cases) the industrial activity, the flow and the Q.O.D. (Author)

  18. Safety

    International Nuclear Information System (INIS)

    2001-01-01

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

  19. Heat transfer calculations for the High Flux Isotope Reactor (HFIR). Technical specifications: bases for safety limits and limiting safety system settings

    International Nuclear Information System (INIS)

    Sims, T.M.; Swanks, J.H.

    1977-09-01

    Heat transfer analyses, in support of the preparation of the HFIR technical specifications, were made to establish the bases for the safety limits and limiting safety system settings applicable to the HFIR. The results of these analyses, along with the detailed bases, are presented

  20. Grit-mediated frictional ignition of a polymer-bonded explosive during oblique impacts: Probability calculations for safety engineering

    International Nuclear Information System (INIS)

    Heatwole, Eric; Parker, Gary; Holmes, Matt; Dickson, Peter

    2015-01-01

    Frictional heating of high-melting-point grit particles during oblique impacts of consolidated explosives is considered to be the major source of ignition in accidents involving dropped explosives. It has been shown in other work that the lower temperature melting point of two frictionally interacting surfaces will cap the maximum temperature reached, which provides a simple way to mitigate the danger in facilities by implementing surfaces with melting points below the ignition temperature of the explosive. However, a recent series of skid testing experiments has shown that ignition can occur on low-melting-point surfaces with a high concentration of grit particles, most likely due to a grit–grit collision mechanism. For risk-based safety engineering purposes, the authors present a method to estimate the probability of grit contact and/or grit–grit collision during an oblique impact. These expressions are applied to potentially high-consequence oblique impact scenarios in order to give the probability of striking one or more grit particles (for high-melting-point surfaces), or the probability of one or more grit–grit collisions occurring (for low-melting-point surfaces). The probability is dependent on a variety of factors, many of which can be controlled for mitigation to achieve acceptable risk levels for safe explosives handling operations. - Highlights: • Unexpectedly, grit-mediated ignition of a PBX occurred on low-melting point surfaces. • On high-melting surfaces frictional heating is due to a grit–surface interaction. • For low-melting point surfaces the heating mechanism is grit–grit collisions. • A method for estimating the probability of ignition is presented for both surfaces

  1. SSI's independent consequence calculations in support of the regulatory review of the SR-Can safety assessment

    International Nuclear Information System (INIS)

    Shulan Xu; Dverstorp, Bjoern; Woerman, Anders; Marklund, Lars; Klos, Richard; Shaw, George

    2008-03-01

    With the publication of the SR-Can report at the end of 2006, Swedish Nuclear Fuel and Waste Management Co (SKB) have presented a complete assessment of long-term safety for a KBS-3 repository. The SR-Can project demonstrates progress in SKB's capabilities in respect of the methodology for assessment of long-term safety in support of a licence application for a final repository. According to SKB's plans, applications to construct a geological repository will be submitted in 2009, supported by post-closure safety assessments. Project CLIMB (Catchment LInked Models of radiological effects in the Biosphere) was instituted in 2004 to provide SSI with an independent modelling capability when reviewing SKB's assessments. Modelling in CLIMB covers all aspects of performance assessment (PA) from nearfield releases to radiological consequences in the surface environment. This review of SR-Can provides the first opportunity to apply the models and to compare the CLIMB approach with developments at SKB. The aim of the independent calculations is to investigate key aspects of the PA models and so to better understand the assessment methodology used by SKB. Independent modelling allows critical review issues to be addressed by the application of alternative models and assumptions. Three reviews are undertaken here: - Reproduction of selected cases from SR-Can in order to demonstrate an adequate understanding of the PA model from details given in the SR-Can documentation. - Alternative conceptualisation of radionuclide transport and accumulation in the surface system. Two modelling approaches have been used: GEMA (the Generic Ecosystem Modelling Approach) is a traditional compartmental model similar to that used by SKB in SR-Can but with additional functionality and flexibility. The second approach takes continuous transport models to investigate contaminant migration through the Quaternary deposits into the surface drainage system. - The final strand of the CLIMB investigation

  2. SSI's independent consequence calculations in support of the regulatory review of the SR-Can safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Shulan Xu; Dverstorp, Bjoern (Swedish Radiation Protection Authority, Stockholm (Sweden)); Woerman, Anders; Marklund, Lars (Royal Institute of Technology (KTH), Stockholm (SE)); Klos, Richard (Aleksandria Sciences, Sheffield (GB)); Shaw, George (Univ. of Nottingham (GB))

    2008-03-15

    With the publication of the SR-Can report at the end of 2006, Swedish Nuclear Fuel and Waste Management Co (SKB) have presented a complete assessment of long-term safety for a KBS-3 repository. The SR-Can project demonstrates progress in SKB's capabilities in respect of the methodology for assessment of long-term safety in support of a licence application for a final repository. According to SKB's plans, applications to construct a geological repository will be submitted in 2009, supported by post-closure safety assessments. Project CLIMB (Catchment LInked Models of radiological effects in the Biosphere) was instituted in 2004 to provide SSI with an independent modelling capability when reviewing SKB's assessments. Modelling in CLIMB covers all aspects of performance assessment (PA) from nearfield releases to radiological consequences in the surface environment. This review of SR-Can provides the first opportunity to apply the models and to compare the CLIMB approach with developments at SKB. The aim of the independent calculations is to investigate key aspects of the PA models and so to better understand the assessment methodology used by SKB. Independent modelling allows critical review issues to be addressed by the application of alternative models and assumptions. Three reviews are undertaken here: - Reproduction of selected cases from SR-Can in order to demonstrate an adequate understanding of the PA model from details given in the SR-Can documentation. - Alternative conceptualisation of radionuclide transport and accumulation in the surface system. Two modelling approaches have been used: GEMA (the Generic Ecosystem Modelling Approach) is a traditional compartmental model similar to that used by SKB in SR-Can but with additional functionality and flexibility. The second approach takes continuous transport models to investigate contaminant migration through the Quaternary deposits into the surface drainage system. - The final strand of the CLIMB

  3. Models used in the SFR1 SAR-08 and KBS-3H safety assessments for calculation of C-14 doses

    International Nuclear Information System (INIS)

    Avila, R.; Proehl, G.

    2008-03-01

    This report presents a set of simplified models for assessment of human exposures resulting from potential underground releases of C-14. These models were used in the SFR1 SAR08 and KBS-3H safety assessments. The proposed models can be used to assess continuous, as well as pulse-like C-14 releases, to various types of biosphere objects: forest ecosystems, agricultural lands, sea basins and lakes. It is also possible to make assessments of exposures resulting from the use of contaminated fresh waters, for example from an impacted well, for irrigation of vegetables. Models are also proposed for scenarios where lakes and sea basins are transformed into terrestrial objects due to land rise, filling of lakes and other natural or human induced processes. The exposure pathways considered in dose calculations with the models are: ingestion of contaminated food and water for both terrestrial and aquatic ecosystems, inhalation of contaminated air for terrestrial ecosystems. The exposure by external irradiation is not considered, as C-14 is a pure low energy beta emitter. The report provides an overview of the behaviour of C-14 in the environment, including an outline of the conceptual assumptions implicit in the proposed models. The proposed models are based on the so-called specific activity approach, which has been recommended by the UNSCEAR and the IAEA for assessment of doses resulting from C-14 releases to the environment from nuclear installations. The equations for estimation of the C-14 specific activities in environmental compartments have been derived from a combination of several realistic and conservative assumptions, which are documented and justified in the report. The models can be used in safety assessments of geological repositories of radioactive waste, to carry out cautious, but still not over conservative dose estimations, which can be compared with regulatory dose constrains. Comparative studies with the models indicate that the worse case situations

  4. 49 CFR 244.15 - Subjects to be addressed in a Safety Integration Plan not involving an amalgamation of operations.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Subjects to be addressed in a Safety Integration... Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION REGULATIONS ON SAFETY INTEGRATION PLANS GOVERNING RAILROAD CONSOLIDATIONS, MERGERS, AND...

  5. A Hybrid Artificial Reputation Model Involving Interaction Trust, Witness Information and the Trust Model to Calculate the Trust Value of Service Providers

    Directory of Open Access Journals (Sweden)

    Gurdeep Singh Ransi

    2014-02-01

    Full Text Available Agent interaction in a community, such as the online buyer-seller scenario, is often uncertain, as when an agent comes in contact with other agents they initially know nothing about each other. Currently, many reputation models are developed that help service consumers select better service providers. Reputation models also help agents to make a decision on who they should trust and transact with in the future. These reputation models are either built on interaction trust that involves direct experience as a source of information or they are built upon witness information also known as word-of-mouth that involves the reports provided by others. Neither the interaction trust nor the witness information models alone succeed in such uncertain interactions. In this paper we propose a hybrid reputation model involving both interaction trust and witness information to address the shortcomings of existing reputation models when taken separately. A sample simulation is built to setup buyer-seller services and uncertain interactions. Experiments reveal that the hybrid approach leads to better selection of trustworthy agents where consumers select more reputable service providers, eventually helping consumers obtain more gains. Furthermore, the trust model developed is used in calculating trust values of service providers.

  6. Comparative Study of some Parameters reported in the Safety Analysis Report of TRIGA MARK II Research reactor with Calculations

    International Nuclear Information System (INIS)

    Chakrobortty, T.K.; Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

    1997-06-01

    An attempt has been made to investigate some of the parametric results reported in the safety Analysis Report (SAR) with the theoretical analysis carried out by different computer codes and data bases. Different neutronics, thermal hydraulics and safety parameters such as core criticality and burnup lifetime, power peaking factor, prompt negative temperature coefficient, neutron flux, pulse characteristics, steady state and transient behaviors of the TRIGA reactor were analyzed. The investigated results were found to be in fairly good agreement with the values reported in the SAR. 12 refs., 14 figs., 1 table (Author)

  7. Stakeholder Involvement in nuclear issues. INSAG-20. A report by the International Nuclear Safety Advisory Group (Russian Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    Many of the world's nuclear power plants were constructed long ago without much public involvement in the associated decision making. It is anticipated, however, that a variety of stakeholders will seek participation in such decisions now as the nuclear option is being revisited in many places. Accidents at Three Mile Island and Chernobyl, among other places, have served to arouse public concern. The development of 'here-and-now' media capabilities has created an awareness that may not have previously existed. Improvements in educational systems and the development of the Internet have made technical information and expertise available to individuals and locations that were previously without them. In addition, consideration of the environmental impacts of various energy strategies has moved to the fore. INSAG has concluded that the expectations of stakeholders of a right to participate in energy decisions are something that the nuclear community must address. Decisions regarding such matters as the siting and construction of a nuclear power plant are no longer largely the domain of a closed community of technical experts and utility executives. Today, the concerns and expectations of all manner of persons and organizations - from the local farmer to the international financial institution - must be considered. This report is intended for use by all stakeholders in the nuclear community - national regulatory authorities, nuclear power plant designers and operators, public interest organizations and individuals, the media and, not to be forgotten, local and national populations. INSAG's fundamental conclusion is that all stakeholders with an interest in nuclear decisions should be provided with an opportunity for full and effective participation in them. With this right, however, come certain obligations on all sides for openness, candour and civility. INSAG is hopeful that this report will help define the interests and roles of the stakeholders

  8. The calculated research of a fuel rod thermo technical safety at various laws of energy liberation and temperature changes

    International Nuclear Information System (INIS)

    Azarov, S.I.

    1995-01-01

    Different questions of the thermo technical safety and efficiency at various thermal loads in steady-state and transitional operating conditions are discussed. Resources up to heat exchange crisis at impulse, step and harmonical action when changing of coolant temperature and energy liberation in fuel are shown

  9. New innovative educational method to prevent accidents involving young road users (aged 15-24 – European Road Safety Tunes

    Directory of Open Access Journals (Sweden)

    Jankowska-Karpa Dagmara

    2017-01-01

    Full Text Available The article presents a new teaching method designed to improve road safety among young road users. Developed under “European Road Safety Tunes”, this international project was cofunded by EU DG MOVE. Its main aim is to improve road safety and minimize the number of road accidents, injuries and fatalities among road users who are 15-24 years old. The Safety Tunes method contains a series of workshops addressed to young vocational school students: cyclists, moped and motor riders and car drivers. The workshops incorporate peer and emotive education, and delivery of road safety related messages through different types of artistic forms. The topics tackled during class address awareness of possible risks and risk-behaviour, prevention of distraction and reduction in young fatalities and serious injuries on the road. All actions within the project are evaluated, both in terms of the impact of the workshops on students’ attitudes towards road safety problems and in terms of process assessment.

  10. A Neural Circuit Mechanism for the Involvements of Dopamine in Effort-Related Choices: Decay of Learned Values, Secondary Effects of Depletion, and Calculation of Temporal Difference Error

    Science.gov (United States)

    2018-01-01

    Abstract Dopamine has been suggested to be crucially involved in effort-related choices. Key findings are that dopamine depletion (i) changed preference for a high-cost, large-reward option to a low-cost, small-reward option, (ii) but not when the large-reward option was also low-cost or the small-reward option gave no reward, (iii) while increasing the latency in all the cases but only transiently, and (iv) that antagonism of either dopamine D1 or D2 receptors also specifically impaired selection of the high-cost, large-reward option. The underlying neural circuit mechanisms remain unclear. Here we show that findings i–iii can be explained by the dopaminergic representation of temporal-difference reward-prediction error (TD-RPE), whose mechanisms have now become clarified, if (1) the synaptic strengths storing the values of actions mildly decay in time and (2) the obtained-reward-representing excitatory input to dopamine neurons increases after dopamine depletion. The former is potentially caused by background neural activity–induced weak synaptic plasticity, and the latter is assumed to occur through post-depletion increase of neural activity in the pedunculopontine nucleus, where neurons representing obtained reward exist and presumably send excitatory projections to dopamine neurons. We further show that finding iv, which is nontrivial given the suggested distinct functions of the D1 and D2 corticostriatal pathways, can also be explained if we additionally assume a proposed mechanism of TD-RPE calculation, in which the D1 and D2 pathways encode the values of actions with a temporal difference. These results suggest a possible circuit mechanism for the involvements of dopamine in effort-related choices and, simultaneously, provide implications for the mechanisms of TD-RPE calculation. PMID:29468191

  11. The Atmospherically Important Reaction of Hydroxyl Radicals with Methyl Nitrate: A Theoretical Study Involving the Calculation of Reaction Mechanisms, Enthalpies, Activation Energies, and Rate Coefficients.

    Science.gov (United States)

    Ng, Maggie; Mok, Daniel K W; Lee, Edmond P F; Dyke, John M

    2017-09-07

    A theoretical study, involving the calculation of reaction enthalpies, activation energies, mechanisms, and rate coefficients, was made of the reaction of hydroxyl radicals with methyl nitrate, an important process for methyl nitrate removal in the earth's atmosphere. Four reaction channels were considered: formation of H 2 O + CH 2 ONO 2 , CH 3 OOH + NO 2 , CH 3 OH + NO 3 , and CH 3 O + HNO 3 . For all channels, geometry optimization and frequency calculations were performed at the M06-2X/6-31+G** level, while relative energies were improved at the UCCSD(T*)-F12/CBS level. The major channel is found to be the H abstraction channel, to give the products H 2 O + CH 2 ONO 2 . The reaction enthalpy (ΔH 298 K RX ) of this channel is computed as -17.90 kcal mol -1 . Although the other reaction channels are also exothermic, their reaction barriers are high (>24 kcal mol -1 ), and therefore these reactions do not contribute to the overall rate coefficient in the temperature range considered (200-400 K). Pathways via three transition states were identified for the H abstraction channel. Rate coefficients were calculated for these pathways at various levels of variational transition state theory including tunneling. The results obtained are used to distinguish between two sets of experimental rate coefficients, measured in the temperature range of 200-400 K, one of which is approximately an order of magnitude greater than the other. This comparison, as well as the temperature dependence of the computed rate coefficients, shows that the lower experimental values are favored. The implications of the results to atmospheric chemistry are discussed.

  12. Application of Probability Calculations to the Study of the Permissible Step and Touch Potentials to Ensure Personnel Safety

    International Nuclear Information System (INIS)

    Eisawy, E.A.

    2011-01-01

    The aim of this paper is to develop a practical method to evaluate the actual step and touch potential distributions in order to determine the risk of failure of the grounding system. The failure probability, indicating the safety level of the grounding system, is related to both applied (stress) and withstand (strength) step or touch potentials. The probability distributions of the applied step and touch potentials as well as the corresponding withstand step and touch potentials which represent the capability of the human body to resist stress potentials are presented. These two distributions are used to evaluate the failure probability of the grounding system which denotes the probability that the applied potential exceeds the withstand potential. The method is accomplished in considering the resistance of the human body, the foot contact resistance and the fault clearing time as an independent random variables, rather than fixed values as treated in the previous analysis in determining the safety requirements for a given grounding system

  13. 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...

  14. 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...

  15. [Systemic approach to ecologic safety at objects with radiation jeopardy, involved into localization of low and medium radioactive waste].

    Science.gov (United States)

    Veselov, E I

    2011-01-01

    The article deals with specifying systemic approach to ecologic safety of objects with radiation jeopardy. The authors presented stages of work and algorithm of decisions on preserving reliability of storage for radiation jeopardy waste. Findings are that providing ecologic safety can cover 3 approaches: complete exemption of radiation jeopardy waste, removal of more dangerous waste from present buildings and increasing reliability of prolonged localization of radiation jeopardy waste at the initial place. The systemic approach presented could be realized at various radiation jeopardy objects.

  16. Organization of public authorities in France for the event of an incident or accident involving nuclear safety: Simulation of a nuclear crisis

    International Nuclear Information System (INIS)

    Cartigny, J.; Majorel, Y.

    1986-01-01

    The French nuclear safety regulations lay down the action to be taken in the event of an incident or accident involving the types of radiological hazard that could arise in a nuclear installation or during the transport of radioactive material. The organization established for this purpose is designed to ensure that the technical measures taken by the authorities responsible for nuclear safety, radiation protection, public order and public safety are fully effective. The Interministerial Nuclear Safety Committee (Comite interministeriel de la securite nucleaire), which reports to the Prime Minister, co-ordinates the measures taken by the public authorities. The public authorities and the operators together organize exercises designed to verify the whole complex of measures foreseen in the event of an incident or accident. These exercises, which have been carried out in a systematic manner in France for some years, are based on scenarios which are as realistic as possible and enable the following objectives to be achieved: (1) analysis of the crisis apparatus (ORSECRAD plans, individual intervention plans, information conventions); (2) uncovering gaps or inadequacies; (3) arrangements for interchange of information between the various participants whose responsibilities involve them in the emergency; and (4) allowance for the information requirements of the media and the population. The information drawn from these exercises enables the various procedures to be improved step by step. (author)

  17. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  18. 49 CFR Appendix B to Part 553 - Statement of Policy: Rulemakings Involving the Assessment of the Functional Equivalence of Safety...

    Science.gov (United States)

    2010-10-01

    ... conditions, driver demographics, driver behavior, occupant behavior (e.g., level of safety belt use), road... FMVSS. One reason for conservatism is that differences from vehicle model to vehicle model and... available for a comparison of two standards. Often there is an abundance of one type of data and little or...

  19. Calculation of reactivity for safety in nuclear reactors; Calculo de la reactividad para seguridad en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Suescun D, D. [Universidad Surcolombiana, Av. Pastrana Borrero - Carrera 1, Neiva, Huila (Colombia); Rojas A, O., E-mail: daniel.suescun@usco.edu.co [Universidad Popular Autonoma del Estado de Puebla, Av. 9 Pte 1908, Barrio de Santiago, 72410 Puebla (Mexico)

    2017-09-15

    The measurement of reactivity is a function of time and its calculation results from the variation in nuclear power from the inverse equation of punctual kinetics. This equation is a differential integral, where the term of the integral conserves the historical power and the differential part is directly related to the period of the reactor. In practice, in a nuclear plant, sensors are required to record the signals. For example, the movements of the control rods that cause the fluctuations of nuclear power over time commonly generate signals with noise, an event that makes difficult to estimate the reactivity. Thus is necessary and very useful to build digital reactivity meters in real time, since allows a reactor to be operated with greater security. The calculation of the reactivity is carried out using punctual kinetics, especially the concentration of delayed neutron precursors. In this work we present a new way to reduce the fluctuations in the calculation of the reactivity, for the high precision we propose the generalization of the predictor and corrector of the Adams-Bashforth-Moulton (ABM) method of order 4 to solve numerically the equations of the point kinetics for the calculation of the reactivity, without using the power history, due to the nature of the equations of the punctual kinetics, the modifiers of the different predictors are used to increase the accuracy in the approximation obtained accompanied by the filter known as Savitzky-Golay (Sg), allow to reduce the fluctuations of reactivity. It is known that the Sg filter softens and does not attenuate the nuclear power regardless of its shape, guarantees to reduce noise levels up to σ = 0.01, with a calculation time step of σ = 0.01, s. This formulation uses a polynomial approximation of Gram, with a degree d = 2, to calculate the convolution coefficients by means of an analytical formula that is implemented computationally and avoids problems of bad conditioning, caused by the inversion of a

  20. Reliability of quay walls using finite element analysiscalibration of partial safety factors in quay wall design by probabilistic plaxis calculations

    OpenAIRE

    Wolters, H.J.; Bakker, K.J.; De Gijt, J.G.

    2013-01-01

    During the last two years, CUR committee 183 has worked on the upgrade of the Dutch Quay Walls handbook (CUR 211), which is to be published in 2013. Two of the main elements that are considered in this new edition are the addition of Finite Element analysis (FEM) as a method for design, comparable to the description in the Handbook Sheet-Pile Structures (CUR 166), and the calibration of partial safety factors design with FEM.With respect to the actuality of this update it must be remembered t...

  1. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  2. Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyun; Kim, Young Ae; Oh, Seung Jong; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

  3. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

  4. Scientific and technical conference Thermophysical experimental and calculating and theoretical studies to justify characteristics and safety of fast reactors. Thermophysics-2012. Book of abstracts

    International Nuclear Information System (INIS)

    Kalyakin, S.G.; Kukharchuk, O.F.; Sorokin, A.P.

    2012-01-01

    The collection includes abstracts of reports of scientific and technical conference Thermophysics-2012 which has taken place on October 24-26, 2012 in Obninsk. In abstracts the following questions are considered: experimental and calculating and theoretical studies of thermal hydraulics of liquid-metal cooled fast reactors to justify their characteristics and safety; physico-chemical processes in the systems with liquid-metal coolants (LMC); physico-chemical characteristics and thermophysical properties of LMC; development of models, computational methods and calculational codes for simulating processes of of hydrodynamics, heat and mass transfer, including impurities mass transfer in the systems with LMC; methods and means for control of composition and condition of LMC in fast reactor circuits on impurities and purification from them; apparatuses, equipment and technological processes at the work with LMC taking into account the ecology, including fast reactors decommissioning; measuring techniques, sensors and devices for experimental studies of heat and mass transfer in the systems with LMC [ru

  5. The importance of Probabilistic Safety Assessment in the careful study of risks involved to new nuclear power plant projects

    International Nuclear Information System (INIS)

    Mata, Jônatas F.C. da; Mesquita, Amir Z.

    2017-01-01

    The Fukushima Daiichi nuclear accident in Japan in 2011 has raised public fears about the actual safety of nuclear power plants in several countries. The response to this concern by government agencies and private companies has been objective and pragmatic in order to guarantee best practices in the design, construction, operation and decommissioning phases of nuclear reactors. In countries where the nucleo-electric matrix is consolidated, such as the United States, France and the United Kingdom, the safety assessment is carried out considering deterministic and probabilistic criteria. In the licensing stages of new projects, it is necessary to analyze and simulate the behavior of the nuclear power plant, when subjected to conditions that can lead to sequences of accidents. Each initiator event is studied and simulated through computational models, which allow the description and estimation of possible physical phenomena occurring in nuclear reactors. Probabilistic Safety Assessment (PSA) is fundamental in this process, as it studies in depth the sequences of events that can lead to the fusion of the nucleus of the nuclear reactor. Such sequences should be quantified in terms of probability of occurrence and your possible consequences, and organized through techniques such as Fault Tree Analysis and Event Tree Analysis. For these simulations, specialized computer codes for each type of phenomenon should be used, as well as databases based on experience gained in the operation of similar nuclear reactors. The present work will describe, in an objective way, the procedures for the realization of PSA and its applicability to the assurance of the operational reliability of the nuclear reactors, as well as a brief comparative between the approaches used in some countries traditionally users of thermonuclear energy and Brazil. By means of this analysis, it can be concluded that nuclear power is increasingly reliable and safe, being able to provide the necessary

  6. The importance of Probabilistic Safety Assessment in the careful study of risks involved to new nuclear power plant projects

    Energy Technology Data Exchange (ETDEWEB)

    Mata, Jônatas F.C. da, E-mail: jonatasfmata@yahoo.com.br [Universidade do Estado de Minas Gerais (UEMG), João Monlevade, MG (Brazil); Mesquita, Amir Z., E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The Fukushima Daiichi nuclear accident in Japan in 2011 has raised public fears about the actual safety of nuclear power plants in several countries. The response to this concern by government agencies and private companies has been objective and pragmatic in order to guarantee best practices in the design, construction, operation and decommissioning phases of nuclear reactors. In countries where the nucleo-electric matrix is consolidated, such as the United States, France and the United Kingdom, the safety assessment is carried out considering deterministic and probabilistic criteria. In the licensing stages of new projects, it is necessary to analyze and simulate the behavior of the nuclear power plant, when subjected to conditions that can lead to sequences of accidents. Each initiator event is studied and simulated through computational models, which allow the description and estimation of possible physical phenomena occurring in nuclear reactors. Probabilistic Safety Assessment (PSA) is fundamental in this process, as it studies in depth the sequences of events that can lead to the fusion of the nucleus of the nuclear reactor. Such sequences should be quantified in terms of probability of occurrence and your possible consequences, and organized through techniques such as Fault Tree Analysis and Event Tree Analysis. For these simulations, specialized computer codes for each type of phenomenon should be used, as well as databases based on experience gained in the operation of similar nuclear reactors. The present work will describe, in an objective way, the procedures for the realization of PSA and its applicability to the assurance of the operational reliability of the nuclear reactors, as well as a brief comparative between the approaches used in some countries traditionally users of thermonuclear energy and Brazil. By means of this analysis, it can be concluded that nuclear power is increasingly reliable and safe, being able to provide the necessary

  7. Insight into "Calculated Risk": An Application to the Prioritization of Emerging Infectious Diseases for Blood Transfusion Safety.

    Science.gov (United States)

    Neslo, R E J; Oei, W; Janssen, M P

    2017-09-01

    Increasing identification of transmissions of emerging infectious diseases (EIDs) by blood transfusion raised the question which of these EIDs poses the highest risk to blood safety. For a number of the EIDs that are perceived to be a threat to blood safety, evidence on actual disease or transmission characteristics is lacking, which might render measures against such EIDs disputable. On the other hand, the fact that we call them "emerging" implies almost by definition that we are uncertain about at least some of their characteristics. So what is the relative importance of various disease and transmission characteristics, and how are these influenced by the degree of uncertainty associated with their actual values? We identified the likelihood of transmission by blood transfusion, the presence of an asymptomatic phase of infection, prevalence of infection, and the disease impact as the main characteristics of the perceived risk of disease transmission by blood transfusion. A group of experts in the field of infectious diseases and blood transfusion ranked sets of (hypothetical) diseases with varying degrees of uncertainty associated with their disease characteristics, and used probabilistic inversion to obtain probability distributions for the weight of each of these risk characteristics. These distribution weights can be used to rank both existing and newly emerging infectious diseases with (partially) known characteristics. Analyses show that in case there is a lack of data concerning disease characteristics, it is the uncertainty concerning the asymptomatic phase and the disease impact that are the most important drivers of the perceived risk. On the other hand, if disease characteristics are well established, it is the prevalence of infection and the transmissibility of the disease by blood transfusion that will drive the perceived risk. The risk prioritization model derived provides an easy to obtain and rational expert assessment of the relative importance of

  8. Exploring the theory, barriers and enablers for patient and public involvement across health, social care and patient safety: a protocol for a systematic review of reviews.

    Science.gov (United States)

    Ocloo, Josephine; Garfield, Sarah; Dawson, Shoba; Dean Franklin, Bryony

    2017-10-24

    The emergence of patient and public involvement (PPI) in healthcare in the UK can be traced as far back as the 1970s. More recently, campaigns by harmed patients and their relatives have emerged as a result of clinical failings in the NHS, challenging paternalistic healthcare, which have led to a new focus on PPI in quality and safety, nationally and internationally. Evidence suggests that PPI within patient safety is often atheoretical and located within a biomedical discourse. This review will explore the literature on PPI across patient safety, healthcare and social care to identify theory, barriers and enablers that can be used to develop PPI in patient safety. Systematic searches of three electronic bibliographic databases will be conducted, using both MeSH and free-text terms to identify empirical literature published from database inception to May 2017. The screening process will involve input from at least two researchers and any disagreement will be resolved through discussion with a third reviewer. Initial inclusion and exclusion criteria have been developed and will be refined iteratively throughout the process. Data extraction from included articles will be conducted by at least two researchers using a data extraction form. Extracted information will be analysed using a narrative review approach, which synthesises data using a descriptive method. No ethical approval is required for this review as no empirical data were collected. We believe that the findings and recommendations from this review will be particularly relevant for an audience of academics and policymakers. The findings will, therefore, be written up and disseminated in international peer-reviewed journals and academic conferences with a health focus. They will also be disseminated to leading health policy organisations in the NHS, such as NHS England and NHS Improvement and national policy bodies such as the Health Foundation. © Article author(s) (or their employer(s) unless otherwise

  9. Challenges in ensuring radiological safety and nuclear forensic for malicious acts involving nuclear and other radioactive material

    International Nuclear Information System (INIS)

    Sharma, Ranjit; Chatterjee, M.K.; Singh, Rajvir; Pradeepkumar, K.S.

    2010-01-01

    Nuclear and other radioactive materials may get smuggled into the country aimed at malicious acts. Radioactive material detected accidentally or during inspection at the entry points/national borders may indicate illicit trafficking for the purpose of nuclear/radiological terrorism. As country requires prevention and preparedness for response to these malicious acts, nuclear forensic techniques are to be developed incorporating radiological safety aspects. Nuclear forensics helps in determining the origin, intended use, legal owner and the smuggled route etc. by using fingerprinting as well as comparison with reference data. The suggested sequence of methods for analysis of radioactive material/samples will be radiological assessment, physical characterization, traditional forensic analysis, isotope analysis along with elemental/chemical analysis

  10. Condition-based fault tree analysis (CBFTA): A new method for improved fault tree analysis (FTA), reliability and safety calculations

    International Nuclear Information System (INIS)

    Shalev, Dan M.; Tiran, Joseph

    2007-01-01

    Condition-based maintenance methods have changed systems reliability in general and individual systems in particular. Yet, this change does not affect system reliability analysis. System fault tree analysis (FTA) is performed during the design phase. It uses components failure rates derived from available sources as handbooks, etc. Condition-based fault tree analysis (CBFTA) starts with the known FTA. Condition monitoring (CM) methods applied to systems (e.g. vibration analysis, oil analysis, electric current analysis, bearing CM, electric motor CM, and so forth) are used to determine updated failure rate values of sensitive components. The CBFTA method accepts updated failure rates and applies them to the FTA. The CBFTA recalculates periodically the top event (TE) failure rate (λ TE ) thus determining the probability of system failure and the probability of successful system operation-i.e. the system's reliability. FTA is a tool for enhancing system reliability during the design stages. But, it has disadvantages, mainly it does not relate to a specific system undergoing maintenance. CBFTA is tool for updating reliability values of a specific system and for calculating the residual life according to the system's monitored conditions. Using CBFTA, the original FTA is ameliorated to a practical tool for use during the system's field life phase, not just during system design phase. This paper describes the CBFTA method and its advantages are demonstrated by an example

  11. Survey data for the application to Japan of international ideas on safety of works involving radiation exposure

    International Nuclear Information System (INIS)

    1979-01-01

    In order to apply ICRP Publication 27 to Japan, various concerned data in the nation were collected and analyzed. The data are the following: (1) for the respective industries, the number of deaths, age distribution of deaths, and frequency of injuries with seriousness and occupational diseases; and (2) for industries involving radiation exposure, the average reduction of life span due to radiation-induced deaths, and bodily, genetic and pregnancy effects of radiation exposure. (Mori, K.)

  12. Safety and efficacy of high-dose melphalan and auto-SCT in patients with AL amyloidosis and cardiac involvement.

    Science.gov (United States)

    Girnius, S; Seldin, D C; Meier-Ewert, H K; Sloan, J M; Quillen, K; Ruberg, F L; Berk, J L; Doros, G; Sanchorawala, V

    2014-03-01

    In Ig light chain (AL) amyloidosis, cardiac involvement is associated with worse prognosis and increased treatment-related complications. In this retrospective cohort study, we assessed survival, hematologic and cardiac responses to high-dose melphalan and auto-SCT (HDM/SCT) in patients with AL amyloidosis and cardiac involvement, stratified by cardiac biomarkers brain natriuretic peptide and Troponin I, analogous to the Mayo cardiac staging. Forty-seven patients underwent HDM/SCT based upon functional measures; six patients had modified cardiac stage I disease, seventeen had modified cardiac stage II disease and twenty-four had modified cardiac stage III disease. Treatment-related mortality was 4% for all patients and 8% for patients with stage III disease. Three-year survival was 88% and EFS was 47%; these did not differ by stage. By intention-to-treat analysis, 27% of patients achieved a hematologic complete response and 32% a very good partial response, of whom 70 and 45%, respectively, have not required additional therapy at 36 months. Cardiac response was achieved in 53% of patients. We conclude that with appropriate patient selection and a risk-adapted treatment approach, HDM/SCT is safe and effective in patients with AL amyloidosis and cardiac involvement.

  13. Reliability Calculations

    DEFF Research Database (Denmark)

    Petersen, Kurt Erling

    1986-01-01

    Risk and reliability analysis is increasingly being used in evaluations of plant safety and plant reliability. The analysis can be performed either during the design process or during the operation time, with the purpose to improve the safety or the reliability. Due to plant complexity and safety...... and availability requirements, sophisticated tools, which are flexible and efficient, are needed. Such tools have been developed in the last 20 years and they have to be continuously refined to meet the growing requirements. Two different areas of application were analysed. In structural reliability probabilistic...... approaches have been introduced in some cases for the calculation of the reliability of structures or components. A new computer program has been developed based upon numerical integration in several variables. In systems reliability Monte Carlo simulation programs are used especially in analysis of very...

  14. Applying the Plan-Do-Study-Act (PDSA) approach to a large pragmatic study involving safety net clinics.

    Science.gov (United States)

    Coury, Jennifer; Schneider, Jennifer L; Rivelli, Jennifer S; Petrik, Amanda F; Seibel, Evelyn; D'Agostini, Brieshon; Taplin, Stephen H; Green, Beverly B; Coronado, Gloria D

    2017-06-19

    The Plan-Do-Study-Act (PDSA) cycle is a commonly used improvement process in health care settings, although its documented use in pragmatic clinical research is rare. A recent pragmatic clinical research study, called the Strategies and Opportunities to STOP Colon Cancer in Priority Populations (STOP CRC), used this process to optimize the research implementation of an automated colon cancer screening outreach program in intervention clinics. We describe the process of using this PDSA approach, the selection of PDSA topics by clinic leaders, and project leaders' reactions to using PDSA in pragmatic research. STOP CRC is a cluster-randomized pragmatic study that aims to test the effectiveness of a direct-mail fecal immunochemical testing (FIT) program involving eight Federally Qualified Health Centers in Oregon and California. We and a practice improvement specialist trained in the PDSA process delivered structured presentations to leaders of these centers; the presentations addressed how to apply the PDSA process to improve implementation of a mailed outreach program offering colorectal cancer screening through FIT tests. Center leaders submitted PDSA plans and delivered reports via webinar at quarterly meetings of the project's advisory board. Project staff conducted one-on-one, 45-min interviews with project leads from each health center to assess the reaction to and value of the PDSA process in supporting the implementation of STOP CRC. Clinic-selected PDSA activities included refining the intervention staffing model, improving outreach materials, and changing workflow steps. Common benefits of using PDSA cycles in pragmatic research were that it provided a structure for staff to focus on improving the program and it allowed staff to test the change they wanted to see. A commonly reported challenge was measuring the success of the PDSA process with the available electronic medical record tools. Understanding how the PDSA process can be applied to pragmatic

  15. Reliability calculations

    International Nuclear Information System (INIS)

    Petersen, K.E.

    1986-03-01

    Risk and reliability analysis is increasingly being used in evaluations of plant safety and plant reliability. The analysis can be performed either during the design process or during the operation time, with the purpose to improve the safety or the reliability. Due to plant complexity and safety and availability requirements, sophisticated tools, which are flexible and efficient, are needed. Such tools have been developed in the last 20 years and they have to be continuously refined to meet the growing requirements. Two different areas of application were analysed. In structural reliability probabilistic approaches have been introduced in some cases for the calculation of the reliability of structures or components. A new computer program has been developed based upon numerical integration in several variables. In systems reliability Monte Carlo simulation programs are used especially in analysis of very complex systems. In order to increase the applicability of the programs variance reduction techniques can be applied to speed up the calculation process. Variance reduction techniques have been studied and procedures for implementation of importance sampling are suggested. (author)

  16. Factors associated with regulatory action involving investigation of illnesses associated with Shiga toxin-producing Escherichia coli in products regulated by the Food Safety and Inspection Service.

    Science.gov (United States)

    Green, Alice L; Seys, Scott; Douris, Aphrodite; Levine, Jeoff; Robertson, Kis

    2014-07-01

    We described characteristics of the Escherichia coli O157 and Escherichia coli non-O157 illness investigations conducted by the United States Department of Agriculture's Food Safety and Inspection Service (FSIS) during the 5-year period from 2006 through 2010. We created a multivariable logistic regression model to determine characteristics of these investigations that were associated with FSIS regulatory action, which was defined as having occurred if a product recall occurred or if FSIS personnel performed an environmental health assessment (Food Safety Assessment) at the implicated establishment. During this period, FSIS took regulatory action in 38 of 88 (43%) investigations. Illness investigations in which FoodNet states were involved were more likely to result in regulatory action. Illness investigations in which state and local traceback, or FSIS traceback occurred were more likely to result in regulatory action. Reasons for lack of action included evidence of cross-contamination after the product left a regulated establishment, delayed notification, lack of epidemiological information, and insufficient product information.

  17. System safety education focused on flight safety

    Science.gov (United States)

    Holt, E.

    1971-01-01

    The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.

  18. 33 CFR 96.320 - What is involved to complete a safety management audit and when is it required to be completed?

    Science.gov (United States)

    2010-07-01

    ... Safety Management (ISM) Code by Administrations. (3) Make sure the audit is carried out by a team of... safety management audit and when is it required to be completed? 96.320 Section 96.320 Navigation and... SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS How Will Safety Management Systems Be...

  19. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  20. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

    International Nuclear Information System (INIS)

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  1. Efficacy and safety of nerve growth factor for the treatment of neurological diseases:a meta-analysis of 64 randomized controlled trials involving 6,297 patients

    Institute of Scientific and Technical Information of China (English)

    Meng Zhao; Xiao-yan Li; Chun-ying Xu; Li-ping Zou

    2015-01-01

    OBJECTIVE:China is the only country where nerve growth factor is approved for large-scale use as a clinical medicine. More than 10 years ago, in 2003, nerve growth factor injection was listed as a national drug. The goal of this article is to evaluate comprehensively the efifcacy and safety of nerve growth factor for the treatment of neurological diseases. DATA RETRIEVAL:A computer-based retrieval was performed from six databases, including the Cochrane Library, PubMed, EMBASE, Sino Med, CNKI, and the VIP database, searching from the clinical establishment of nerve growth factor for treatment until December 31, 2013. The key words for the searches were “nerve growth factor, randomized controlled trials” in Chinese and in English. DATA SELECTION:Inclusion criteria: any study published in English or Chinese referring to randomized controlled trials of nerve growth factor; patients with neurological diseases such as peripheral nerve injury, central nerve injury, cranial neuropathy, and nervous system infections;patients older than 7 years; similar research methods and outcomes assessing symptoms; and measurement of nerve conduction velocities. The meta-analysis was conducted using Review Manager 5.2.3 software. MAIN OUTCOME MEASURES:The total effective rate, the incidence of adverse effects, and the nerve conduction velocity were recorded for each study. RESULTS:Sixty-four studies involving 6,297 patients with neurological diseases were included. The total effective rate in the group treated with nerve growth factor was significantly higher than that in the control group (P < 0.0001,RR: 1.35, 95%CI: 1.30–1.40). The average nerve conduction velocity in the nerve growth factor group was signiifcantly higher than that in the control group (P < 0.00001,MD: 4.59 m/s, 95%CI: 4.12–5.06). The incidence of pain or sclero-ma at the injection site in the nerve growth factor group was also higher than that in the control group (P < 0.00001,RR: 6.30, 95%CI: 3.53–11

  2. Can patient involvement improve patient safety? A cluster randomised control trial of the Patient Reporting and Action for a Safe Environment (PRASE) intervention.

    Science.gov (United States)

    Lawton, Rebecca; O'Hara, Jane Kathryn; Sheard, Laura; Armitage, Gerry; Cocks, Kim; Buckley, Hannah; Corbacho, Belen; Reynolds, Caroline; Marsh, Claire; Moore, Sally; Watt, Ian; Wright, John

    2017-08-01

    To evaluate the efficacy of the Patient Reporting and Action for a Safe Environment intervention. A multicentre cluster randomised controlled trial. Clusters were 33 hospital wards within five hospitals in the UK. All patients able to give informed consent were eligible to take part. Wards were allocated to the intervention or control condition. The ward-level intervention comprised two tools: (1) a questionnaire that asked patients about factors contributing to safety (patient measure of safety (PMOS)) and (2) a proforma for patients to report both safety concerns and positive experiences (patient incident reporting tool). Feedback was considered in multidisciplinary action planning meetings. Primary outcomes were routinely collected ward-level harm-free care (HFC) scores and patient-level feedback on safety (PMOS). Intervention uptake and retention of wards was 100% and patient participation was high (86%). We found no significant effect of the intervention on any outcomes at 6 or 12 months. However, for new harms (ie, those for which the wards were directly accountable) intervention wards did show greater, though non-significant, improvement compared with control wards. Analyses also indicated that improvements were largest for wards that showed the greatest compliance with the intervention. Adherence to the intervention, particularly the implementation of action plans, was poor. Patient safety outcomes may represent too blunt a measure. Patients are willing to provide feedback about the safety of their care. However, we were unable to demonstrate any overall effect of this intervention on either measure of patient safety and therefore cannot recommend this intervention for wider uptake. Findings indicate promise for increasing HFC where wards implement ≥75% of the intervention components. ISRCTN07689702; pre-results. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

  3. Models of radionuclide distribution in the biosphere for radioactive waste storage safety assessment, collection of data and calculation of the biosphere dose conversion factors. Research report

    International Nuclear Information System (INIS)

    Landa, Jiri

    2008-12-01

    The core of the report is structured as follows: The biosphere dose conversion factor (BDCF); Foreign approaches (Sweden - SKB, USA - YMP, BIOPROTA); Definition and conversion factors for activity; Effective dose rate calculation (ingestion, inhalation, external irradiation); Analysis of the activity of the surface compartment, i.e. soil; Basic conceptual models of ecosystems; BDCF calculation/determination; and Systemization of the literature. (P.A.)

  4. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  5. Fisheries and aquaculture industries involvement to control product health and quality safety to satisfy consumer-driven objectives on retail markets in Europe

    International Nuclear Information System (INIS)

    Roth, Eva; Rosenthal, Harald

    2006-01-01

    Over the past years the export of agricultural and fishery products from developing countries has substantially increased to markets within the OECD. Retailers and importers are expanding their international operations to meet consumer demands for year-round delivery of products. Moreover, consumers have become increasingly concerned about the safety of food, including those derived from aquatic resources [FAO/NACA/WHO Joint Study Group, 1999. Report food safety issues associated with products from aquaculture. WHO Technical Report Series No 883: VII, pp. 1-55]. Governments and leading businesses are responding by imposing new safety regulations and standards to the international food system (e.g. HACCP, EUREP-GAP), product liability and labeling [Reilly, A., Howgate, P., Kaeferstein, F., 1997. Safety hazards and the application of HACCP in aquaculture. In: Proceedings of the Second International Conference on Fish Inspection and Quality Control: A Global Focus, Arlington, VA, 19-24 May 1996. Technomic Publishing, Lancaster, PA, pp. 353-373]. Initial concerns for imports of aquacultural products from developing to industrialized countries focussed on bacterial contamination [Buras, N. 1993. Microbial safety of produce from wastewater-fed aquaculture. In: Pullin, R.V.C., Rosenthal, H., MacLean, J.L.(Eds.), Proceedings of ICLARM Conferences, vol. 31, pp. 285-295]. Today, if trade opportunities are to be maintained, these countries must adapt to a full array of regulations and standards. This paper describes four scenarios in aquaculture and fishing product trade between developing countries and countries in the European Union

  6. Fisheries and aquaculture industries involvement to control product health and quality safety to satisfy consumer-driven objectives on retail markets in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Roth, Eva [University of South Denmark, Department of Environmental and Business Economics, Niels Bohrs vej 9, DK-6700 Esbjerg (Denmark); Institute for Marine Research, University Kiel, Duesternbrooker Weg 20, 24105 Kiel (Germany); Rosenthal, Harald [University of South Denmark, Department of Environmental and Business Economics, Niels Bohrs vej 9, DK-6700 Esbjerg (Denmark); Institute for Marine Research, University Kiel, Duesternbrooker Weg 20, 24105 Kiel (Germany)

    2006-07-01

    Over the past years the export of agricultural and fishery products from developing countries has substantially increased to markets within the OECD. Retailers and importers are expanding their international operations to meet consumer demands for year-round delivery of products. Moreover, consumers have become increasingly concerned about the safety of food, including those derived from aquatic resources [FAO/NACA/WHO Joint Study Group, 1999. Report food safety issues associated with products from aquaculture. WHO Technical Report Series No 883: VII, pp. 1-55]. Governments and leading businesses are responding by imposing new safety regulations and standards to the international food system (e.g. HACCP, EUREP-GAP), product liability and labeling [Reilly, A., Howgate, P., Kaeferstein, F., 1997. Safety hazards and the application of HACCP in aquaculture. In: Proceedings of the Second International Conference on Fish Inspection and Quality Control: A Global Focus, Arlington, VA, 19-24 May 1996. Technomic Publishing, Lancaster, PA, pp. 353-373]. Initial concerns for imports of aquacultural products from developing to industrialized countries focussed on bacterial contamination [Buras, N. 1993. Microbial safety of produce from wastewater-fed aquaculture. In: Pullin, R.V.C., Rosenthal, H., MacLean, J.L.(Eds.), Proceedings of ICLARM Conferences, vol. 31, pp. 285-295]. Today, if trade opportunities are to be maintained, these countries must adapt to a full array of regulations and standards. This paper describes four scenarios in aquaculture and fishing product trade between developing countries and countries in the European Union.

  7. Calculator calculus

    CERN Document Server

    McCarty, George

    1982-01-01

    How THIS BOOK DIFFERS This book is about the calculus. What distinguishes it, however, from other books is that it uses the pocket calculator to illustrate the theory. A computation that requires hours of labor when done by hand with tables is quite inappropriate as an example or exercise in a beginning calculus course. But that same computation can become a delicate illustration of the theory when the student does it in seconds on his calculator. t Furthermore, the student's own personal involvement and easy accomplishment give hi~ reassurance and en­ couragement. The machine is like a microscope, and its magnification is a hundred millionfold. We shall be interested in limits, and no stage of numerical approximation proves anything about the limit. However, the derivative of fex) = 67.SgX, for instance, acquires real meaning when a student first appreciates its values as numbers, as limits of 10 100 1000 t A quick example is 1.1 , 1.01 , 1.001 , •••• Another example is t = 0.1, 0.01, in the functio...

  8. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of excedeence of damage by integrated Safety Analysis Methodology

    International Nuclear Information System (INIS)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-01-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Excedence Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  9. Enhanced safety features of CHASHMA NPP UNIT-2 to encounter selected severe accidents, various challenges involved to prove the adequacy of severe accidents prevention/mitigation measures and to write management guidelines with one possible solution to these challenges

    International Nuclear Information System (INIS)

    Iqbal, Z.; Minhaj, A.

    2007-01-01

    This paper describes enhanced safety features of Chashma Nuclear Power Plant Unit-2 (C-2), a 325 MWe PWR to encounter selected severe accidents and discusses various challenges involved to prove the adequacy of severe accidents encountering measures and to write severe accident management guidelines (SAMGs) in compliance with the recently introduced national regulations based on the new IAEA nuclear safety standards. C-2 is being built by China National Nuclear Corporation (CNNC) for Pakistan Atomic Energy Commission (PAEC). Its twin, Unit-1 (C-1) also a 325 MWe PWR, was commissioned in 2000. Nuclear power safety with reference to severe accidents should be treated as a global issue and therefore the developed countries should include the people of developing countries in nuclear power industry's various severe accidents based research and development programs. The implementation of this idea may also deliver few other useful and mutually beneficial byproducts. (author)

  10. Safety in numbers 4: The relationship between exposure to authentic and didactic environments and nursing students' learning of medication dosage calculation problem solving knowledge and skills.

    Science.gov (United States)

    Weeks, Keith W; Clochesy, John M; Hutton, B Meriel; Moseley, Laurie

    2013-03-01

    Advancing the art and science of education practice requires a robust evaluation of the relationship between students' exposure to learning and assessment environments and the development of their cognitive competence (knowing that and why) and functional competence (know-how and skills). Healthcare education translation research requires specific education technology assessments and evaluations that consist of quantitative analyses of empirical data and qualitative evaluations of the lived student experience of the education journey and schemata construction (Weeks et al., 2013a). This paper focuses on the outcomes of UK PhD and USA post-doctorate experimental research. We evaluated the relationship between exposure to traditional didactic methods of education, prototypes of an authentic medication dosage calculation problem-solving (MDC-PS) environment and nursing students' construction of conceptual and calculation competence in medication dosage calculation problem-solving skills. Empirical outcomes from both UK and USA programmes of research identified highly significant differences in the construction of conceptual and calculation competence in MDC-PS following exposure to the authentic learning environment to that following exposure to traditional didactic transmission methods of education (p students exposure to authentic learning environments is an essential first step in the development of conceptual and calculation competence and relevant schemata construction (internal representations of the relationship between the features of authentic dosage problems and calculation functions); and how authentic environments more ably support all cognitive (learning) styles in mathematics than traditional didactic methods of education. Functional competence evaluations are addressed in Macdonald et al. (2013) and Weeks et al. (2013e). Copyright © 2012. Published by Elsevier Ltd.

  11. Fisheries and aquaculture industries involvement to control product health and quality safety to satisfy consumer-driven objectives on retail markets in Europe.

    Science.gov (United States)

    Roth, Eva; Rosenthal, Harald

    2006-01-01

    Over the past years the export of agricultural and fishery products from developing countries has substantially increased to markets within the OECD. Retailers and importers are expanding their international operations to meet consumer demands for year-round delivery of products. Moreover, consumers have become increasingly concerned about the safety of food, including those derived from aquatic resources [FAO/NACA/WHO Joint Study Group, 1999. Report food safety issues associated with products from aquaculture. WHO Technical Report Series No 883: VII, pp. 1-55]. Governments and leading businesses are responding by imposing new safety regulations and standards to the international food system (e.g. HACCP, EUREP-GAP), product liability and labeling [Reilly, A., Howgate, P., Käferstein, F., 1997. Safety hazards and the application of HACCP in aquaculture. In: Proceedings of the Second International Conference on Fish Inspection and Quality Control: A Global Focus, Arlington, VA, 19-24 May 1996. Technomic Publishing, Lancaster, PA, pp. 353-373]. Initial concerns for imports of aquacultural products from developing to industrialized countries focussed on bacterial contamination [Buras, N. 1993. Microbial safety of produce from wastewater-fed aquaculture. In: Pullin, R.V.C., Rosenthal, H., MacLean, J.L.(Eds.), Proceedings of ICLARM Conferences, vol. 31, pp. 285-295]. Today, if trade opportunities are to be maintained, these countries must adapt to a full array of regulations and standards. This paper describes four scenarios in aquaculture and fishing product trade between developing countries and countries in the European Union.

  12. Safety in Cryogenics – Safety device sizing

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    The calculation is separated in three operations: o The estimation of the loads arriving on the component to protect, o The calculation of the mass flow to evacuate, o And the sizing of the safety device.

  13. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

  14. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model

  15. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements; Calculos neutronicos, termo-hidraulicos e de seguranca de um dispositivo para Irradiacao de miniplacas (DIM) de elementos combustiveis tipo dispersao

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges

    2010-07-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a Miniplate Irradiation Device (MID) to be placed in the IEA-R1 reactor core. The irradiation device is used to receive miniplates of U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}- Al dispersion fuels, LEU type (19.75 % {sup 235}U) with uranium densities of, respectively, 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 235}U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and 2DB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  16. Safety Evaluation of Chinese Medicine Injections with a Cell Imaging-Based Multiparametric Assay Revealed a Critical Involvement of Mitochondrial Function in Hepatotoxicity

    Directory of Open Access Journals (Sweden)

    Meng Wang

    2015-01-01

    Full Text Available The safety of herbal medicine products has been a widespread concern due to their complex chemical nature and lack of proper evaluation methods. We have adapted a sensitive and reproducible multiparametric cell-based high-content analysis assay to evaluate the hepatic-safety of four Chinese medicine injections and validated it with classical animal-based toxicity assays. Our results suggested that the reported hepatotoxicity by one of the drugs, Fufangkushen injection, could be attributed at least in part to the interference of mitochondrial function in human HepG2 cells by some of its constituents. This method should be useful for both preclinical screen in a drug discovery program and postclinical evaluation of herbal medicine preparations.

  17. National Institute for Occupational Safety and Health Oversight: OMB Involvement in VDT Study. Hearing before the Subcommittee on Health and Safety of the Committee on Education and Labor. House of Representatives, Ninety-Ninth Congress, Second Session (June 4, 1986).

    Science.gov (United States)

    Congress of the U.S., Washington, DC. House Committee on Education and Labor.

    This hearing addressed the issue of whether the delays in producing a proposed National Institute for Occupational and Safety Health (NIOSH) study on the possible health hazards associated with video display terminals (VDTs) are due to concerns about scientific methodology or unwarranted interference by the Office of Management and Budget (OMB).…

  18. Safety in numbers 7: Veni, vidi, duci: a grounded theory evaluation of nursing students' medication dosage calculation problem-solving schemata construction.

    Science.gov (United States)

    Weeks, Keith W; Higginson, Ray; Clochesy, John M; Coben, Diana

    2013-03-01

    This paper evaluates nursing students' transition through schemata construction and competence development in medication dosage calculation problem-solving (MDC-PS). We advance a grounded theory from interview data that reflects the experiences and perceptions of two groups of undergraduate pre-registration nursing students: eight students exposed to a prototype authentic MDC-PS environment and didactic transmission methods of education and 15 final year students exposed to the safeMedicate authentic MDC-PS environment. We advance a theory of how classroom-based 'chalk and talk' didactic transmission environments offered multiple barriers to accurate MDC-PS schemata construction among novice students. While conversely it was universally perceived by all students that authentic learning and assessment environments enabled MDC-PS schemata construction through facilitating: 'seeing' the authentic features of medication dosage problems; context-based and situational learning; learning within a scaffolded environment that supported construction of cognitive links between the concrete world of clinical MDC-PS and the abstract world of mathematics; and confidence-building in their cognitive and functional competence ability. Drawing on the principle of veni, vidi, duci (I came, I saw, I calculated), we combined the two sets of evaluations to offer a grounded theoretical basis for schemata construction and competence development within this critical domain of professional practice. Copyright © 2012 Elsevier Ltd. All rights reserved.

  19. Activities for Calculators.

    Science.gov (United States)

    Hiatt, Arthur A.

    1987-01-01

    Ten activities that give learners in grades 5-8 a chance to explore mathematics with calculators are provided. The activity cards involve such topics as odd addends, magic squares, strange projects, and conjecturing rules. (MNS)

  20. Declination Calculator

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Declination is calculated using the current International Geomagnetic Reference Field (IGRF) model. Declination is calculated using the current World Magnetic Model...

  1. A EU simulation platform for nuclear reactor safety: multi-scale and multi-physics calculations, sensitivity and uncertainty analysis (NURESIM project)

    International Nuclear Information System (INIS)

    Chauliac, Christian; Bestion, Dominique; Crouzet, Nicolas; Aragones, Jose-Maria; Cacuci, Dan Gabriel; Weiss, Frank-Peter; Zimmermann, Martin A.

    2010-01-01

    The NURESIM project, the numerical simulation platform, is developed in the frame of the NURISP European Collaborative Project (FP7), which includes 22 organizations from 14 European countries. NURESIM intends to be a reference platform providing high quality software tools, physical models, generic functions and assessment results. The NURESIM platform provides an accurate representation of the physical phenomena by promoting and incorporating the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling. It includes multi-scale and multi-physics features, especially for coupling core physics and thermal-hydraulics models for reactor safety. Easy coupling of the different codes and solvers is provided through the use of a common data structure and generic functions (e.g., for interpolation between non-conforming meshes). More generally, the platform includes generic pre-processing, post-processing and supervision functions through the open-source SALOME software, in order to make the codes more user-friendly. The platform also provides the informatics environment for testing and comparing different codes. The contribution summarizes the achievements and ongoing developments of the simulation platform in core physics, thermal-hydraulics, multi-physics, uncertainties and code integration

  2. Nuclear Safety

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1989-01-01

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  3. Craniocaudal Safety Margin Calculation Based on Interfractional Changes in Tumor Motion in Lung SBRT Assessed With an EPID in Cine Mode

    International Nuclear Information System (INIS)

    Ueda, Yoshihiro; Miyazaki, Masayoshi; Nishiyama, Kinji; Suzuki, Osamu; Tsujii, Katsutomo; Miyagi, Ken

    2012-01-01

    Purpose: To evaluate setup error and interfractional changes in tumor motion magnitude using an electric portal imaging device in cine mode (EPID cine) during the course of stereotactic body radiation therapy (SBRT) for non–small-cell lung cancer (NSCLC) and to calculate margins to compensate for these variations. Materials and Methods: Subjects were 28 patients with Stage I NSCLC who underwent SBRT. Respiratory-correlated four-dimensional computed tomography (4D-CT) at simulation was binned into 10 respiratory phases, which provided average intensity projection CT data sets (AIP). On 4D-CT, peak-to-peak motion of the tumor (M-4DCT) in the craniocaudal direction was assessed and the tumor center (mean tumor position [MTP]) of the AIP (MTP-4DCT) was determined. At treatment, the tumor on cone beam CT was registered to that on AIP for patient setup. During three sessions of irradiation, peak-to-peak motion of the tumor (M-cine) and the mean tumor position (MTP-cine) were obtained using EPID cine and in-house software. Based on changes in tumor motion magnitude (∆M) and patient setup error (∆MTP), defined as differences between M-4DCT and M-cine and between MTP-4DCT and MTP-cine, a margin to compensate for these variations was calculated with Stroom’s formula. Results: The means (±standard deviation: SD) of M-4DCT and M-cine were 3.1 (±3.4) and 4.0 (±3.6) mm, respectively. The means (±SD) of ∆M and ∆MTP were 0.9 (±1.3) and 0.2 (±2.4) mm, respectively. Internal target volume-planning target volume (ITV-PTV) margins to compensate for ∆M, ∆MTP, and both combined were 3.7, 5.2, and 6.4 mm, respectively. Conclusion: EPID cine is a useful modality for assessing interfractional variations of tumor motion. The ITV-PTV margins to compensate for these variations can be calculated.

  4. Impact of preoperative calculation of nephron volume loss on future of partial nephrectomy techniques; planning a strategic roadmap for improving functional preservation and securing oncological safety.

    Science.gov (United States)

    Rha, Koon H; Abdel Raheem, Ali; Park, Sung Y; Kim, Kwang H; Kim, Hyung J; Koo, Kyo C; Choi, Young D; Jung, Byung H; Lee, Sang K; Lee, Won K; Krishnan, Jayram; Shin, Tae Y; Cho, Jin-Seon

    2017-11-01

    To assess the correlation of the resected and ischaemic volume (RAIV), which is a preoperatively calculated volume of nephron loss, with the amount of postoperative renal function (PRF) decline after minimally invasive partial nephrectomy (PN) in a multi-institutional dataset. We identified 348 patients from March 2005 to December 2013 at six institutions. Data on all cases of laparoscopic (n = 85) and robot-assisted PN (n = 263) performed were retrospectively gathered. Univariable and multivariable linear regression analyses were used to identify the associations between various time points of PRF and the RAIV, as a continuous variable. The mean (sd) RAIV was 24.2 (29.2) cm 3 . The mean preoperative estimated glomerular filtration rate (eGFR) and the eGFRs at postoperative day 1, 6 and 36 months after PN were 91.0 and 76.8, 80.2 and 87.7 mL/min/1.73 m 2 , respectively. In multivariable linear regression analysis, the amount of decline in PRF at follow-up was significantly correlated with the RAIV (β 0.261, 0.165, 0.260 at postoperative day 1, 6 and 36 months after PN, respectively). This study has the limitation of its retrospective nature. Preoperatively calculated RAIV significantly correlates with the amount of decline in PRF during long-term follow-up. The RAIV could lead our research to the level of prediction of the amount of PRF decline after PN and thus would be appropriate for assessing the technical advantages of emerging techniques. © 2017 The Authors BJU International © 2017 BJU International Published by John Wiley & Sons Ltd.

  5. Analysis and evaluation of critical experiments for validation of neutron transport calculations

    International Nuclear Information System (INIS)

    Bazzana, S.; Blaumann, H; Marquez Damian, J.I

    2009-01-01

    The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook. [es

  6. Analysis of a calculation method for the determination of the value of safety or control bars; Analisis de un metodo de calculo para la determinacion del valor de barras de seguridad o control

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F; Torres A, C; Filio L, C [ININ, Gcia. de Reactores, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1982-09-15

    Due to the control or safety bars in a nuclear reactor are constituted by strongly absorbent materials, the Diffusion Theory like tool for the calculation of bar values is not directly applicable, should it use the Transport Theory. However the speed and economy of the Diffusion codes for the reactors calculation, those make attractiveness and by this reason its are used in the determination of characteristic parameters and even in the determination of bar values, not without before to make some theoretical developments that allow to make applicable this theory. The application of the Diffusion Theory in strongly absorbent media is based on the use of some effective cross sections distinct from the real ones obtained when imposing the reason that among the flow and it gradient in the external surface of such media (control element in general, bar type or flagstone) be similar to the one obtained using Transport Theory in all the control region (multiplicative and absorbent media) with those real cross sections. The effective cross sections were obtained of the Leopard-NUMICE cell code which has incorporate the respective calculation theory of effective cross sections. Later these constants its were used in the bidimensional diffusion code Exterminator-II, simulating in it, the distribution of safety or control bars. From the cell code its were also obtained the respective constants of the homogeneous fuel cell. The results as soon as those obtained bar values of the diffusion code, its were compared with some experimental results obtained in the R{phi} Swedish reactor of natural uranium and heavy water. In this work an analysis of the bar value of one of them, trying to determine the applicability of the method is made. (Author)

  7. Motor Vehicle Safety

    Science.gov (United States)

    ... these crashes is one part of motor vehicle safety. Here are some things you can do to ... speed or drive aggressively Don't drive impaired Safety also involves being aware of others. Share the ...

  8. Radiation safety assessment and development of environmental radiation monitoring technology; standardization of input parameters for the calculation of annual dose from routine releases from commercial reactor effluents

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I. H.; Cho, D.; Youn, S. H.; Kim, H. S.; Lee, S. J.; Ahn, H. K. [Soonchunhyang University, Ahsan (Korea)

    2002-04-01

    This research is to develop a standard methodology for determining the input parameters that impose a substantial impact on radiation doses of residential individuals in the vicinity of four nuclear power plants in Korea. We have selected critical nuclei, pathways and organs related to the human exposure via simulated estimation with K-DOSE 60 based on the updated ICRP-60 and sensitivity analyses. From the results, we found that 1) the critical nuclides were found to be {sup 3}H, {sup 133}Xe, {sup 60}Co for Kori plants and {sup 14}C, {sup 41}Ar for Wolsong plants. The most critical pathway was 'vegetable intake' for adults and 'milk intake' for infants. However, there was no preference in the effective organs, and 2) sensitivity analyses showed that the chemical composition in a nuclide much more influenced upon the radiation dose than any other input parameters such as food intake, radiation discharge, and transfer/concentration coefficients by more than 102 factor. The effect of transfer/concentration coefficients on the radiation dose was negligible. All input parameters showed highly estimated correlation with the radiation dose, approximated to 1.0, except for food intake in Wolsong power plant (partial correlation coefficient (PCC)=0.877). Consequently, we suggest that a prediction model or scenarios for food intake reflecting the current living trend and a formal publications including details of chemical components in the critical nuclei from each plant are needed. Also, standardized domestic values of the parameters used in the calculation must replace the values of the existed or default-set imported factors via properly designed experiments and/or modelling such as transport of liquid discharge in waters nearby the plants, exposure tests on corps and plants so on. 4 figs., 576 tabs. (Author)

  9. Researches in nuclear safety

    International Nuclear Information System (INIS)

    Souchet, Y.

    2009-01-01

    This article comprises three parts: 1 - some general considerations aiming at explaining the main motivations of safety researches, and at briefly presenting the important role of some organisations in the international conciliation, and the most common approach used in safety researches (analytical experiments, calculation codes, global experiments); 2 - an overview of some of the main safety problems that are the object of worldwide research programs (natural disasters, industrial disasters, criticality, human and organisational factors, fuel behaviour in accidental situation, serious accidents: core meltdown, corium spreading, failure of the confinement building, radioactive releases). Considering the huge number of research topics, this part cannot be exhaustive and many topics are not approached; 3 - the presentation of two research programs addressing very different problems: the evaluation of accidental releases in the case of a serious accident (behaviour of iodine and B 4 C, air infiltration, fission products release) and the propagation of a fire in a facility (PRISME program). These two programs belong to an international framework involving several partners from countries involved in nuclear energy usage. (J.S.)

  10. Radiation safety

    International Nuclear Information System (INIS)

    Jain, Priyanka

    2014-01-01

    The use of radiation sources is a privilege; in order to retain the privilege, all persons who use sources of radiation must follow policies and procedures for their safe and legal use. The purpose of this poster is to describe the policies and procedures of the Radiation Protection Program. Specific conditions of radiation safety require the establishment of peer committees to evaluate proposals for the use of radionuclides, the appointment of a radiation safety officer, and the implementation of a radiation safety program. In addition, the University and Medical Centre administrations have determined that the use of radiation producing machines and non-ionizing radiation sources shall be included in the radiation safety program. These Radiation Safety policies are intended to ensure that such use is in accordance with applicable State and Federal regulations and accepted standards as directed towards the protection of health and the minimization of hazard to life or property. It is the policy that all activities involving ionizing radiation or radiation emitting devices be conducted so as to keep hazards from radiation to a minimum. Persons involved in these activities are expected to comply fully with the Canadian Nuclear Safety Act and all it. The risk of prosecution by the Department of Health and Community Services exists if compliance with all applicable legislation is not fulfilled. (author)

  11. Fatigue-related crashes involving express buses in Malaysia: will the proposed policy of banning the early-hour operation reduce fatigue-related crashes and benefit overall road safety?

    Science.gov (United States)

    Mohamed, Norlen; Mohd-Yusoff, Mohammad-Fadhli; Othman, Ilhamah; Zulkipli, Zarir-Hafiz; Osman, Mohd Rasid; Voon, Wong Shaw

    2012-03-01

    Fatigue-related crashes have long been the topic of discussion and study worldwide. The relationship between fatigue-related crashes and time of day is well documented. In Malaysia, the possibility of banning express buses from operating during the early-hours of the morning has emerged as an important consideration for passenger safety. This paper highlights the findings of an impact assessment study. The study was conducted to determine all possible impacts prior to the government making any decision on the proposed banning. This study is an example of a simple and inexpensive approach that may influence future policy-making process. The impact assessment comprised two major steps. The first step involved profiling existing operation scenarios, gathering information on crashes involving public express buses and stakeholders' views. The second step involved a qualitative impact assessment analysis using all information gathered during the profiling stage to describe the possible impacts. Based on the assessment, the move to ban early-hour operations could possibly result in further negative impacts on the overall road safety agenda. These negative impacts may occur if the fundamental issues, such as driving and working hours, and the need for rest and sleep facilities for drivers, are not addressed. In addition, a safer and more accessible public transportation system as an alternative for those who choose to travel at night would be required. The proposed banning of early-hour operations is also not a feasible solution for sustainability of express bus operations in Malaysia, especially for those operating long journeys. The paper concludes by highlighting the need to design a more holistic approach for preventing fatigue-related crashes involving express buses in Malaysia. Copyright © 2011 Elsevier Ltd. All rights reserved.

  12. Thermodynamic Database for the Terrestrial and Planetary Mantle Studies: Where we stand, and some future directions involving experimental studies, numerical protocol for EoS and atomistic calculations (Invited)

    Science.gov (United States)

    Ganguly, J.; Tirone, M.; Sorcar, N.

    2013-12-01

    problem of appropriate combination of binary mixing properties in a multicomponent system, and present simplified a-x formalisms for complex multi-site solutions the full-blown treatment of which requires a body of experimental data that are unlikely to be available in the near future. It appears that Cp and the coefficients of thermal expansion and compression of solids could be calculated quite well form density functional theory (DFT) (e.g. 7), so that the future development of databases should not only consider the available DFT data, but also have active involvement in such studies 'to fill the gap' when reliable data are not available. While the current trend in geodynamic modeling is to use thermodynamic properties in tabulated form, more realistic simulations, which we would try to illustrate by examples, would require real-time thermodynamic calculations for evolving bulk compositions; hence the development of a robust yet simple thermodynamic formulation becomes essential. 1. Fabrichnaya et al. (2004), Springer-verlag; 2. Stixrude and Lithgow-Bertelloni, Geophys. J. Int. (2011) 184, 1180-1213. 3. Holland et al. ( 2013) J. Petrol. 4. Report, Geomaterials Genome Project, March 19-23, 2013, Miami, Florida, NSF Geoinformatics program. 5. Ferreira et al. (1988) Phys Rev B, 37, 10547-10570; 6. Ganguly et al. (1993) Amer Min 78, 583-593 7. Ottonello et a. (2009) PCM 36, 87-106

  13. Shields calculations for teletherapy equipment. Regulatory approach of the National Center of Nuclear Safety; Calculos de blindajes para equipos de teleterapia, enfoque regulatorio del Centro Nacional de Seguridad Nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y. [Centro Nacional de Seguridad Nuclear, Calle 28 No. 504 e/5 y 7 Ave. Miramar, Playa, Ciudad de la Habana (Cuba)]. e-mail: andres@orasen.co.cu

    2006-07-01

    The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)

  14. PROBLEMS OF APPLYING FIXED FORMULAE TO SAFETY CRITERIA AND SITE SELECTION

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. K.

    1963-10-15

    The problem of developing a formula or calculation procedure for that could more-or-less automatically indicate whether or not a nuclear plant would be considered safe at a particular location is discussed. The difficulties and impossibilities of any sach formula for making decisions on siting and safety involving large amounts of money and public safety are considered. (P.C.H.)

  15. CONTAIN calculations

    International Nuclear Information System (INIS)

    Scholtyssek, W.

    1995-01-01

    In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

  16. How safe is the safety paradigm?

    NARCIS (Netherlands)

    O.A. Arah (Onyebuchi); N.S. Klazinga (Niek)

    2004-01-01

    textabstractThis paper reviews safety initiatives in the health systems of the UK, Canada, Australia, and the US. Initiatives to tackle safety shortcomings involve public-private collaborations. Patient safety agencies (to institute learning, action and safety culture), adverse

  17. Current interruption transients calculation

    CERN Document Server

    Peelo, David F

    2014-01-01

    Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,

  18. Playground Safety

    Science.gov (United States)

    ... Prevention Fall Prevention Playground Safety Poisoning Prevention Road Traffic Safety Sports Safety Get Email Updates To receive ... at the Consumer Product Safety Commission’s Playground Safety website . References U.S. Consumer Product Safety Commission. Injuries and ...

  19. Burnout calculation

    International Nuclear Information System (INIS)

    Li, D.

    1980-01-01

    Reviewed is the effect of heat flux of different system parameters on critical density in order to give an initial view on the value of several parameters. A thorough analysis of different equations is carried out to calculate burnout is steam-water flows in uniformly heated tubes, annular, and rectangular channels and rod bundles. Effect of heat flux density distribution and flux twisting on burnout and storage determination according to burnout are commended [ru

  20. The Involved Ostrich

    DEFF Research Database (Denmark)

    Davies, Andrea; Dobscha, Susan; Geiger, Susi

    2008-01-01

    that the transition into parenthood can be difficult for men due to their lack of a physical connection to the pregnancy, a perception that the baby industry is not designed for them, the continuance of male stereotypes in the media, and also the time available to men to become involved in consumption activities......-natal data. Data revealed that men, according to their partner’s perceptions, used consumption as a virtual umbilical cord, although levels of consumption involvement varied from co-involvement for most purchases, to limited involvement, and/or involvement for ‘large’ items, particularly travel systems...... and technical items. This research also revealed that men partook in highly masculinized forms of “nesting,” and in general shunned pregnancy book reading; although some did engage in “research” activities such as searching the internet for product safety information. We conclude from this study...

  1. DOE handbook electrical safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    Electrical Safety Handbook presents the Department of Energy (DOE) safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety guidance and information for DOE installations to effect a reduction or elimination of risks associated with the use of electrical energy. The objectives of this handbook are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.

  2. Electrical safety guidelines

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-01

    The Electrical Safety Guidelines prescribes the DOE safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety standards and guidance for DOE installations in order to affect a reduction or elimination of risks associated with the use of electrical energy. The objectives of these guidelines are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.

  3. Hypervelocity impact cratering calculations

    Science.gov (United States)

    Maxwell, D. E.; Moises, H.

    1971-01-01

    A summary is presented of prediction calculations on the mechanisms involved in hypervelocity impact cratering and response of earth media. Considered are: (1) a one-gram lithium-magnesium alloys impacting basalt normally at 6.4 km/sec, and (2) a large terrestrial impact corresponding to that of Sierra Madera.

  4. Managing Parent Involvement during Crisis

    Science.gov (United States)

    Merriman, Lynette S.

    2008-01-01

    In the wake of 9/11, Hurricane Katrina, and the Virginia Tech shooting tragedy, it is no surprise that concern for students' safety is the primary reason attributed to parents' increased involvement. Parents and university administrators share in their commitment to student safety. However, college and university staff who assume responsibility…

  5. Introduction to 'International Handbook of Criticality Safety Benchmark Experiments'

    International Nuclear Information System (INIS)

    Komuro, Yuichi

    1998-01-01

    The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated year by year by the working group of the project. This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used. The author briefly introduces the informative handbook and would like to encourage Japanese engineers who are in charge of nuclear criticality safety to use the handbook. (author)

  6. Importance of the Electron Correlation and Dispersion Corrections in Calculations Involving Enamines, Hemiaminals, and Aminals. Comparison of B3LYP, M06-2X, MP2, and CCSD Results with Experimental Data.

    Science.gov (United States)

    Castro-Alvarez, Alejandro; Carneros, Héctor; Sánchez, Dani; Vilarrasa, Jaume

    2015-12-18

    While B3LYP, M06-2X, and MP2 calculations predict the ΔG° values for exchange equilibria between enamines and ketones with similar acceptable accuracy, the M06-2X/6-311+G(d,p) and MP2/6-311+G(d,p) methods are required for enamine formation reactions (for example, for enamine 5a, arising from 3-methylbutanal and pyrrolidine). Stronger disagreement was observed when calculated energies of hemiaminals (N,O-acetals) and aminals (N,N-acetals) were compared with experimental equilibrium constants, which are reported here for the first time. Although it is known that the B3LYP method does not provide a good description of the London dispersion forces, while M06-2X and MP2 may overestimate them, it is shown here how large the gaps are and that at least single-point calculations at the CCSD(T)/6-31+G(d) level should be used for these reaction intermediates; CCSD(T)/6-31+G(d) and CCSD(T)/6-311+G(d,p) calculations afford ΔG° values in some cases quite close to MP2/6-311+G(d,p) while in others closer to M06-2X/6-311+G(d,p). The effect of solvents is similarly predicted by the SMD, CPCM, and IEFPCM approaches (with energy differences below 1 kcal/mol).

  7. Hydrogen peroxide safety issues

    International Nuclear Information System (INIS)

    Conner, W.V.

    1993-01-01

    A literature survey was conducted to review the safety issues involved in handling hydrogen peroxide solutions. Most of the information found in the literature is not directly applicable to conditions at the Rocky Flats Plant, but one report describes experimental work conducted previously at Rocky Flats to determine decomposition reaction-rate constants for hydrogen peroxide solutions. Data from this report were used to calculate decomposition half-life times for hydrogen peroxide in solutions containing several decomposition catalysts. The information developed from this survey indicates that hydrogen peroxide will undergo both homogeneous and heterogeneous decomposition. The rate of decomposition is affected by temperature and the presence of catalytic agents. Decomposition of hydrogen peroxide is catalyzed by alkalies, strong acids, platinum group and transition metals, and dissolved salts of transition metals. Depending upon conditions, the consequence of a hydrogen peroxide decomposition can range from slow evolution of oxygen gas to a vapor, phase detonation of hydrogen peroxide vapors

  8. Parental involvement

    Directory of Open Access Journals (Sweden)

    Ezra S Simon

    2005-01-01

    Full Text Available Parent-Teacher Associations and other community groups can play a significant role in helping to establish and run refugee schools; their involvement can also help refugee adults adjust to their changed circumstances.

  9. Nuclear criticality safety parameter evaluation for uranium metallic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

    2013-07-01

    Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

  10. Reactors also involve people

    International Nuclear Information System (INIS)

    Hurt, H.B.

    1975-01-01

    As the nuclear industry develops it is to be hoped that high quality occupational health programs will evolve along with other sound operational procedures and practices. The immediate involvement of occupational health personnel may well afford a safety factor which will minimize the likelihood of either the selection of personnel not adequate for the full responsibilities of their work or the continuation in responsible positions of personnel who develop handicaps of either a physical or mental nature

  11. Ensuring the validity of calculated subcritical limits

    International Nuclear Information System (INIS)

    Clark, H.K.

    1977-01-01

    The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

  12. Hydrogen safety

    International Nuclear Information System (INIS)

    Frazier, W.R.

    1991-01-01

    The NASA experience with hydrogen began in the 1950s when the National Advisory Committee on Aeronautics (NACA) research on rocket fuels was inherited by the newly formed National Aeronautics and Space Administration (NASA). Initial emphasis on the use of hydrogen as a fuel for high-altitude probes, satellites, and aircraft limited the available data on hydrogen hazards to small quantities of hydrogen. NASA began to use hydrogen as the principal liquid propellant for launch vehicles and quickly determined the need for hydrogen safety documentation to support design and operational requirements. The resulting NASA approach to hydrogen safety requires a joint effort by design and safety engineering to address hydrogen hazards and develop procedures for safe operation of equipment and facilities. NASA also determined the need for rigorous training and certification programs for personnel involved with hydrogen use. NASA's current use of hydrogen is mainly for large heavy-lift vehicle propulsion, which necessitates storage of large quantities for fueling space shots and for testing. Future use will involve new applications such as thermal imaging

  13. On the energy scale involved in the metal to insulator transition of quadruple perovskite EuCu3Fe4O12: infrared spectroscopy and ab-initio calculations.

    Science.gov (United States)

    Brière, B; Kalinko, A; Yamada, I; Roy, P; Brubach, J B; Sopracase, R; Zaghrioui, M; Phuoc, V Ta

    2016-06-27

    Optical measurements were carried out by infrared spectroscopy on AA'3B4O12 A-site ordered quadruple perovskite EuCu3Fe4O12 (microscopic sample) as function of temperature. At 240 K (=TMI), EuCu3Fe4O12 undergoes a very abrupt metal to insulator transition, a paramagnetic to antiferromagnetic transition and an isostructural transformation with an abrupt large volume expansion. Above TMI, optical conductivity reveals a bad metal behavior and below TMI, an insulating phase with an optical gap of 125 meV is observed. As temperature is decreased, a large and abrupt spectral weight transfer toward an energy scale larger than 1 eV is detected. Concurrently, electronic structure calculations for both high and low temperature phases were compared to the optical conductivity results giving a precise pattern of the transition. Density of states and computed optical conductivity analysis identified Cu3dxy, Fe3d and O2p orbitals as principal actors of the spectral weight transfer. The present work constitutes a first step to shed light on EuCu3Fe4O12 electronic properties with optical measurements and ab-initio calculations.

  14. Nuclear safety in France

    International Nuclear Information System (INIS)

    Tanguy, P.

    1979-01-01

    A brief description of the main safety aspects of the French nuclear energy programme and of the general safety organization is followed by a discussion on the current thinking in CEA on some important safety issues. As far as methodology is concerned, the use of probabilistic analysis in the licensing procedure is being extensively developed. Reactor safety research is aimed at a better knowledge of the safety margins involved in the present designs of both PWRs and LMFBRs. A greater emphasis should be put during the next years in the safety of the nuclear fuel cycle installations, including waste disposals. Finally, it is suggested that further international cooperation in the field of nuclear safety should be developed in order to insure for all countries the very high safety level which has been achieved up till now. (author)

  15. What price safety. A probabilistic cost-benefit evaluaton of existing engineered safety features

    International Nuclear Information System (INIS)

    O'Donnell, E.P.

    1978-01-01

    The paper provides a method for performing quantitative cost-benefit evaluations for nuclear safety concerns involving accidents of low probability and potentially large consequences. It presents an application of the method to ECCS, containment, emergency power system and hydrogen recombiner system. This evaluation provides a valuable assessment of the relative cost effectiveness of these features in reducing accident risk. It also provides insight into the sensitivity of cost-benefit calculations to the manner in which safety features are sequantially added in design. (author)

  16. Auto Safety

    Science.gov (United States)

    ... Safe Videos for Educators Search English Español Auto Safety KidsHealth / For Parents / Auto Safety What's in this ... by teaching some basic rules. Importance of Child Safety Seats Using a child safety seat (car seat) ...

  17. Safety first

    Energy Technology Data Exchange (ETDEWEB)

    Harvie, W.

    1997-06-01

    Expansion of international business opportunities for Canadian producers and service companies brings with it a dimension almost never considered on home base - security. It was pointed out that once abroad, safety and defence of people and equipment can become significant problems in many parts of the world. The nature of the security risks involved, and how best to deal with them, were discussed. The use of consultants, mostly foreign ones to date, and the kind of assistance they can provide, everything from written reports on the local situation to counter surveillance training, and bodyguard services, have been described. Examples of recent involvements with guerilla groups demanding `revolutionary war taxes`, kidnapping executives for ransom, due diligence investigations of potential partners, and the like, have been provided to illustrate the unique character of the problem, and the constant need for being alert, educated to risks, and being prepared to react to risk situations.

  18. Safety in cardiac surgery

    NARCIS (Netherlands)

    Siregar, S.

    2013-01-01

    The monitoring of safety in cardiac surgery is a complex process, which involves many clinical, practical, methodological and statistical issues. The objective of this thesis was to measure and to compare safety in cardiac surgery in The Netherlands using the Netherlands Association for

  19. Elements of nuclear safety

    CERN Document Server

    Libmann, Jacques

    1996-01-01

    This basically educational book is intended for all involved in nuclear facility safety. It dissects the principles and experiences conducive to the adoption of attitudes compliant with what is now known as "safety culture". This book is accessible to a wide range of readers.

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

    International Nuclear Information System (INIS)

    Vovk, Ivan

    1998-01-01

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

  1. Leadership for safety: industrial experience.

    Science.gov (United States)

    Flin, R; Yule, S

    2004-12-01

    The importance of leadership for effective safety management has been the focus of research attention in industry for a number of years, especially in energy and manufacturing sectors. In contrast, very little research into leadership and safety has been carried out in medical settings. A selective review of the industrial safety literature for leadership research with possible application in health care was undertaken. Emerging findings show the importance of participative, transformational styles for safety performance at all levels of management. Transactional styles with attention to monitoring and reinforcement of workers' safety behaviours have been shown to be effective at the supervisory level. Middle managers need to be involved in safety and foster open communication, while ensuring compliance with safety systems. They should allow supervisors a degree of autonomy for safety initiatives. Senior managers have a prime influence on the organisation's safety culture. They need to continuously demonstrate a visible commitment to safety, best indicated by the time they devote to safety matters.

  2. An international nuclear safety regime

    International Nuclear Information System (INIS)

    Rosen, M.

    1995-01-01

    For all the parties involved with safe use of nuclear energy, the opening for signature of the 'Convention on Nuclear Safety' (signed by 60 countries) and the ongoing work to prepare a 'Convention on Radioactive Waste Safety' are particularly important milestones. 'Convention on Nuclear Safety' is the first legal instrument that directly addresses the safety of nuclear power plants worldwide. The two conventions are only one facet of international cooperation to enhance safety. A review of some cooperative efforts of the past decades, and some key provisions of the new safety conventions, presented in this paper, show how international cooperation is increasing nuclear safety worldwide. The safety philosophy and practices involved with legal framework for the safe use of nuclear power will foster a collective international involvement and commitment. It will be a positive step towards increasing public confidence in nuclear power

  3. The impact of masculinity on safety oversights, safety priority and safety violations in two male-dominated occupations

    DEFF Research Database (Denmark)

    Nielsen, Kent; Hansen, Claus D.; Bloksgaard, Lotte

    2015-01-01

    Background Although men have a higher risk of occupational injuries than women the role of masculinity for organizational safety outcomes has only rarely been the object of research. Aim The current study investigated the association between masculinity and safety oversights, safety priority......-related context factors (safety leadership, commitment of the safety representative, and safety involvement) and three safety-related outcome factors (safety violations, safety oversights and safety priority) were administered twice 12 months apart to Danish ambulance workers (n = 1157) and slaughterhouse workers...

  4. Involving women.

    Science.gov (United States)

    Agbo, J

    1994-01-01

    I am a primary health care (PHC) coordinator working with the May Day Rural project, a local NGO involved in integrated approaches and programs with rural communities in the Ga District of the Greater-Accra region in Ghana. When we talk about the community development approach we must first and foremost recognize that we are talking about women, because in the developing world frequent childbirths mean that her burden of mortality is higher than a man's; her workload is extremely heavy--whether in gardening, farming, other household duties, caring for the sick, or the rearing of children; she has a key role in PHC and community development, because men are always looking for greener pastures elsewhere, leaving the women behind. Women's concerns are critical in most health care projects and women and children are their main beneficiaries. Why not include women in the management team, project design, implementation and evaluation processes? That is what the May Day Rural project is practicing, encouraging women's participation and creating a relationship of trust. full text

  5. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    fuels as well as the applied methodologies. The IRSN proceeds in a relevant and independent assessment of the submitted safety reports. To achieve this goal and maintain over time an independent and relevant assessment capability, the IRSN relies on the excellence of its experts and on state of art techniques and knowledge. The IRSN contributes by its work in key area to cutting edge research and development in order to drive nuclear industry towards making the best use of scientific and technological progress for improving safety, environmental protection and health. To maintain at all times the state of the art knowledge and the operational expertise necessary to deal efficiently with major nuclear accident consequences, the IRSN carries out, on the one hand, its own research and development programs to gain accurate knowledge on still unknown phenomena for safety analysis. On the other hand, the IRSN works out its own scientific calculation methodologies involving industrial calculation chain. Concerning more particularly the 'two-phase flows' thematic, The ISRN must correctly simulate the primary fluid behavior in the reactor in normal operation as well as in accidental situations, to estimate if, in such situations, the core reactor state is fully safe and any safety risk is undergone The research and development programs launched at the ISRN on two-phase flows gather work on advanced thermohydraulic configurations encounter in various reactor states (normal operation or accidental situations), in particular: (i)The estimation of the margin to the critical heat flux in normal operation (DNBR), (ii) The pressurized thermal shock, which is due to mechanical important constraints in the reactor vessel resulting from the injection of a cold fluid in case of emergency cooling (PTS), (iii) The reactivity insertion accident (RIA), (iv) The loss of coolant accident (LOCA), (vi) The accidents in spent-fuel pools and (vii) The severe accident, which could lead to core

  6. Safety Teams: An Approach to Engage Students in Laboratory Safety

    Science.gov (United States)

    Alaimo, Peter J.; Langenhan, Joseph M.; Tanner, Martha J.; Ferrenberg, Scott M.

    2010-01-01

    We developed and implemented a yearlong safety program into our organic chemistry lab courses that aims to enhance student attitudes toward safety and to ensure students learn to recognize, demonstrate, and assess safe laboratory practices. This active, collaborative program involves the use of student "safety teams" and includes…

  7. Calculation of research reactor RA power at uncontrolled reactivity changes

    International Nuclear Information System (INIS)

    Cupac, S.

    1978-01-01

    The safety analysis of research reactor RA involves also the calculation of reactor power at uncontrolled reactivity changes. The corresponding computer code, based on Point Kinetics Model has been made. The short review of method applied for solving kinetic equations is given and several examples illustrating the reactor behaviour at various reactivity changes are presented. The results already obtained are giving rather rough picture of reactor behaviour in considered situations. This is the consequence of using simplified feed back and reactor cooling models, as well as temperature reactivity coefficients, which do not correspond to the actual reactor RA structure (which is now only partly fulfilled with 80% enriched uranium fuel). (author) [sr

  8. Nuclear safety

    International Nuclear Information System (INIS)

    1991-02-01

    This book reviews the accomplishments, operations, and problems faced by the defense Nuclear Facilities Safety Board. Specifically, it discusses the recommendations that the Safety Board made to improve safety and health conditions at the Department of Energy's defense nuclear facilities, problems the Safety Board has encountered in hiring technical staff, and management problems that could affect the Safety Board's independence and credibility

  9. Deep penetration calculations

    International Nuclear Information System (INIS)

    Thompson, W.L.; Deutsch, O.L.; Booth, T.E.

    1980-04-01

    Several Monte Carlo techniques are compared in the transport of neutrons of different source energies through two different deep-penetration problems each with two parts. The first problem involves transmission through a 200-cm concrete slab. The second problem is a 90 0 bent pipe jacketed by concrete. In one case the pipe is void, and in the other it is filled with liquid sodium. Calculations are made with two different Los Alamos Monte Carlo codes: the continuous-energy code MCNP and the multigroup code MCMG

  10. Drug Safety

    Science.gov (United States)

    ... over-the-counter drug. The FDA evaluates the safety of a drug by looking at Side effects ... clinical trials The FDA also monitors a drug's safety after approval. For you, drug safety means buying ...

  11. Nuclear safety

    International Nuclear Information System (INIS)

    Tarride, Bruno

    2015-10-01

    The author proposes an overview of methods and concepts used in the nuclear industry, at the design level as well as at the exploitation level, to ensure an acceptable safety level, notably in the case of nuclear reactors. He first addresses the general objectives of nuclear safety and the notion of acceptable risk: definition and organisation of nuclear safety (relationships between safety authorities and operators), notion of acceptable risk, deterministic safety approach and main safety principles (safety functions and confinement barriers, concept of defence in depth). Then, the author addresses the safety approach at the design level: studies of operational situations, studies of internal and external aggressions, safety report, design principles for important-for-safety systems (failure criterion, redundancy, failure prevention, safety classification). The next part addresses safety during exploitation and general exploitation rules: definition of the operation domain and of its limits, periodic controls and tests, management in case of incidents, accidents or aggressions

  12. Introduction to safety theory

    International Nuclear Information System (INIS)

    Meyna, A.

    1982-01-01

    After a general introduction to safety theory, safety characteristics are defined and quantified. This is followed by a calculation of the safety characteristics of simple, safety-relevant systems in general and in consideration of common-mode errors. The qualitative and quantitative role of human errors is discussed for various models, and a simple man-machine model is developed for investigation of common-mode errors and human error. The main part of the paper deals with safety analysis in complex systems. After a general review, the common inductive and deductive methods of analysis are presented and commented on and their fields of application discussed. Analytical and simulation codes are presented as methods of evaluation for big, complex event trees - i.e. ''hazard trees in the sense of safety engineering (as a subset of safety relevance). After a basic classification and mathematical formulation of Markovian processes, the author shows that these may be used successfully for calculation of safety characteristics if transition rates are constant and if the number of system states is limited. (orig./RW) [de

  13. Weldon Spring dose calculations

    International Nuclear Information System (INIS)

    Dickson, H.W.; Hill, G.S.; Perdue, P.T.

    1978-09-01

    In response to a request by the Oak Ridge Operations (ORO) Office of the Department of Energy (DOE) for assistance to the Department of the Army (DA) on the decommissioning of the Weldon Spring Chemical Plant, the Health and Safety Research Division of the Oak Ridge National Laboratory (ORNL) performed limited dose assessment calculations for that site. Based upon radiological measurements from a number of soil samples analyzed by ORNL and from previously acquired radiological data for the Weldon Spring site, source terms were derived to calculate radiation doses for three specific site scenarios. These three hypothetical scenarios are: a wildlife refuge for hunting, fishing, and general outdoor recreation; a school with 40 hr per week occupancy by students and a custodian; and a truck farm producing fruits, vegetables, meat, and dairy products which may be consumed on site. Radiation doses are reported for each of these scenarios both for measured uranium daughter equilibrium ratios and for assumed secular equilibrium. Doses are lower for the nonequilibrium case

  14. Safety culture

    International Nuclear Information System (INIS)

    Keen, L.J.

    2003-01-01

    Safety culture has become a topic of increasing interest for industry and regulators as issues are raised on safety problems around the world. The keys to safety culture are organizational effectiveness, effective communications, organizational learning, and a culture that encourages the identification and resolution of safety issues. The necessity of a strong safety culture places an onus on all of us to continually question whether the safety measures already in place are sufficient, and are being applied. (author)

  15. Safety goals and safety culture opening plenary. 2. Safety Regulation Implemented by Gosatomnadzor of Russia

    International Nuclear Information System (INIS)

    Gutsalov, A.T.; Bukrinsky, A.M.

    2001-01-01

    more strict than those recommended in the INSAG-3 and INSAG-12 reports, but they correlate with the value of negligible individual risk of 10 -6 , established in 'Radiation Safety Standards' (NRB-99) and consider still a high level of uncertainty in calculation of these probabilities. OPB- 88/97 also defines safety culture and principles of its formation and provision. Gosatomnadzor of Russia is a federal executive authority implementing state safety regulation in nuclear energy use. One of the main activities of Gosatomnadzor of Russia is nuclear and radiation safety regulation in sitting, design, construction, operation, and decommissioning of nuclear facilities. The activities include the following: 1. development and enactment of regulatory documents; 2. licensing of activities at nuclear facilities; 3. state supervision on observing the requirements of federal rules and regulations and license conditions. Gosatomnadzor of Russia strives toward solving the problems of consistent safety improvement of facilities and technologies up to the internationally accepted level, acting within the framework of the existing set of special safety rules and regulations in production and use of nuclear energy. Simultaneously, Gosatomnadzor of Russia develops proposals aimed at the improvement of legislative and regulatory bases of the Russian Federation as well as licensing and inspection procedures and implementing them. The main principles that Gosatomnadzor of Russia follows in its practical activities are openness, publicity, and cooperation with juridical and natural persons, whose activities are regulated with the purpose of achieving safety. This cooperation is accomplished in compliance with the principle of separation of responsibilities. According to this principle, the parties that are involved in activities related to the use of nuclear materials and nuclear energy on one hand, and in the state regulation of nuclear and radiation safety on the other hand, bear

  16. A balancing method for calculating a component raw involving CGF

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.; Kang, D.; Yang, J.E. [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejon (Korea, Republic of)

    2004-07-01

    In this paper, a method called the 'Balancing Method' to derive a component RAW (Risk Achievement Worth) with basic event RAWs including a CCF (Common Cause Failure) RAW is summarized, and compared with the method proposed by the NEI (Nuclear Energy Institute) by mathematically checking the background on which the two methods are based. It is proved that the Balancing Method has a strong mathematically background. While the NEI method significantly underestimates the component RAW and is a little bit ad hoc in handling CCF RAW, the Balancing Method estimates the true component RAW very closely. Validity of the Balancing Method is based on the fact that if an component is out-of-service, it does not mean that the component is non-existent, but integrates the possibility that the component might fail due to CCF. The validity of the Balancing Method is proved by comparing it to the exact component RAW generated from the fault tree model.

  17. A balancing method for calculating a component raw involving CGF

    International Nuclear Information System (INIS)

    Kim, K.; Kang, D.; Yang, J.E.

    2004-01-01

    In this paper, a method called the 'Balancing Method' to derive a component RAW (Risk Achievement Worth) with basic event RAWs including a CCF (Common Cause Failure) RAW is summarized, and compared with the method proposed by the NEI (Nuclear Energy Institute) by mathematically checking the background on which the two methods are based. It is proved that the Balancing Method has a strong mathematically background. While the NEI method significantly underestimates the component RAW and is a little bit ad hoc in handling CCF RAW, the Balancing Method estimates the true component RAW very closely. Validity of the Balancing Method is based on the fact that if an component is out-of-service, it does not mean that the component is non-existent, but integrates the possibility that the component might fail due to CCF. The validity of the Balancing Method is proved by comparing it to the exact component RAW generated from the fault tree model

  18. Problems involved in calculating the probability of rare occurrences

    International Nuclear Information System (INIS)

    Tittes, E.

    1986-01-01

    Also with regard to the characteristics such as occurrence probability or occurrence rate, there are limits which have to be observed, or else probability data and thus the concept of determinable risk itself will lose its practical value. The mathematical models applied for probability assessment are based on data supplied by the insurance companies, reliability experts in the automobile industry, or by planning experts in the field of traffic or information supply. (DG) [de

  19. Quality assurance for software important to safety

    International Nuclear Information System (INIS)

    2000-01-01

    Software applications play an increasingly relevant role in nuclear power plant systems. This is particularly true of software important to safety used in both: calculations for the design, testing and analysis of nuclear reactor systems (design, engineering and analysis software); and monitoring, control and safety functions as an integral part of the reactor systems (monitoring, control and safety system software). Computer technology is advancing at a fast pace, offering new possibilities in nuclear reactor design, construction, commissioning, operation, maintenance and decommissioning. These advances also present new issues which must be considered both by the utility and by the regulatory organization. Refurbishment of ageing instrumentation and control systems in nuclear power plants and new safety related application areas have emerged, with direct (e.g. interfaces with safety systems) and indirect (e.g. operator intervention) implications for safety. Currently, there exist several international standards and guides on quality assurance for software important to safety. However, none of the existing documents provides comprehensive guidance to the developer, manager and regulator during all phases of the software life-cycle. The present publication was developed taking into account the large amount of available documentation, the rapid development of software systems and the need for updated guidance on h ow to do it . It provides information and guidance for defining and implementing quality assurance programmes covering the entire life-cycle of software important to safety. Expected users are managers, performers and assessors from nuclear utilities, regulatory bodies, suppliers and technical support organizations involved with the development and use of software applied in nuclear power plants

  20. Management of safety culture

    International Nuclear Information System (INIS)

    Kavsek, D.

    2004-01-01

    The strengthening of safety culture in an organization has become an increasingly important issue for nuclear industry. A high level of safety performance is essential for business success in intensely competitive global environment. This presentation offers a discussion of some principles and activities used in enhancing safety performance and appropriate safety behaviour at the Krsko NPP. Over the years a number of events have occurred in nuclear industry that have involved problems in human performance. A review of these and other significant events has identified recurring weaknesses in plant safety culture and policy. Focusing attention on the strengthening of relevant processes can help plants avoid similar undesirable events. The policy of the Krsko NPP is that all employees concerned shall constantly be alert to opportunities to reduce risks to the lowest practicable level and to achieve excellence in plant safety. The most important objective is to protect individuals, society and the environment by establishing and maintaining an effective defense against radiological hazard in the nuclear power plant. It is achieved through the use of reliable structures, components, systems, and procedures, as well as plant personnel committed to a strong safety culture. The elements of safety culture include both organizational and individual aspects. Elements commonly included at the organizational level are senior management commitment to safety, organizational effectiveness, effective communication, organizational learning, and a culture that encourages identification and resolution of safety issues. Elements identified at the individual level include personal accountability, a questioning attitude, communication, procedural adherence, etc.(author)

  1. Safety first

    CERN Multimedia

    2012-01-01

    Safety is a priority for CERN. That is a message I conveyed in my New Year’s address and that I reiterated at one of the first Enlarged Directorate meetings of 2012 when I outlined five key safety objectives for the year, designed and implemented according to accepted international standards.   As we move from spring to summer, it’s time to take stock of how we are doing. Objective number one for 2012, which overarches everything else, is to limit the number of incidents in the workplace. That means systematically investigating and acting on every incident that involves work stoppage, along with all the most frequent workplace accidents: falls, trips and slips. The performance indicator we set ourselves is the percentage of investigations and follow-ups completed. Year on year, these figures are rising but we can never be complacent, and must strive to reach and sustain 100% follow-up. The second objective is to improve hazard control, with a focus in 2012 on chemical ha...

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

    International Nuclear Information System (INIS)

    McEwan, C.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  4. Chapter No.4. Safety analyses

    International Nuclear Information System (INIS)

    2002-01-01

    In 2001 the activity in the field of safety analyses was focused on verification of the safety analyses reports for NPP V-2 Bohunice and NPP Mochovce concerning the new profiled fuel and probabilistic safety assessment study for NPP Mochovce. The calculation safety analyses were performed and expert reviews for the internal UJD needs were elaborated. An important part of work was performed also in solving of scientific and technical tasks appointed within bilateral projects of co-operation between UJD and its international partnership organisations as well as within international projects ordered and financed by the European Commission. All these activities served as an independent support for UJD in its deterministic and probabilistic safety assessment of nuclear installations. A special attention was paid to a review of probabilistic safety assessment study of level 1 for NPP Mochovce. The probabilistic safety analysis of NPP related to the full power operation was elaborated in the study and a contribution of the technical and operational improvements to the risk decreasing was quantified. A core damage frequency of the reactor was calculated and the dominant initiating events and accident sequences with the major contribution to the risk were determined. The target of the review was to determine the acceptance of the sources of input information, assumptions, models, data, analyses and obtained results, so that the probabilistic model could give a real picture of the NPP. The review of the study was performed in co-operation of UJD with the IAEA (IPSART mission) as well as with other external organisations, which were not involved in the elaboration of the reviewed document and probabilistic model of NPP. The review was made in accordance with the IAEA guidelines and methodical documents of UJD and US NRC. In the field of calculation safety analyses the UJD activity was focused on the analysis of an operational event, analyses of the selected accident scenarios

  5. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  6. Evaluation of periodic safety status analyses

    International Nuclear Information System (INIS)

    Faber, C.; Staub, G.

    1997-01-01

    In order to carry out the evaluation of safety status analyses by the safety assessor within the periodical safety reviews of nuclear power plants safety goal oriented requirements have been formulated together with complementary evaluation criteria. Their application in an inter-disciplinary coopertion covering the subject areas involved facilitates a complete safety goal oriented assessment of the plant status. The procedure is outlined briefly by an example for the safety goal 'reactivity control' for BWRs. (orig.) [de

  7. MAXimising Involvement in MUltiMorbidity (MAXIMUM) in primary care: protocol for an observation and interview study of patients, GPs and other care providers to identify ways of reducing patient safety failures.

    Science.gov (United States)

    Daker-White, Gavin; Hays, Rebecca; Esmail, Aneez; Minor, Brian; Barlow, Wendy; Brown, Benjamin; Blakeman, Thomas; Bower, Peter

    2014-08-18

    Increasing numbers of older people are living with multiple long-term health conditions but global healthcare systems and clinical guidelines have traditionally focused on the management of single conditions. Having two or more long-term conditions, or 'multimorbidity', is associated with a range of adverse consequences and poor outcomes and could put patients at increased risk of safety failures. Traditionally, most research into patient safety failures has explored hospital or inpatient settings. Much less is known about patient safety failures in primary care. Our core aims are to understand the mechanisms by which multimorbidity leads to safety failures, to explore the different ways in which patients and services respond (or fail to respond), and to identify opportunities for intervention. We plan to undertake an applied ethnographic study of patients with multimorbidity. Patients' interactions and environments, relevant to their healthcare, will be studied through observations, diary methods and semistructured interviews. A framework, based on previous studies, will be used to organise the collection and analysis of field notes, observations and other qualitative data. This framework includes the domains: access breakdowns, communication breakdowns, continuity of care errors, relationship breakdowns and technical errors. Ethical approval was received from the National Health Service Research Ethics Committee for Wales. An individual case study approach is likely to be most fruitful for exploring the mechanisms by which multimorbidity leads to safety failures. A longitudinal and multiperspective approach will allow for the constant comparison of patient, carer and healthcare worker expectations and experiences related to the provision, integration and management of complex care. This data will be used to explore ways of engaging patients and carers more in their own care using shared decision-making, patient empowerment or other relevant models. Published by

  8. Closure and Sealing Design Calculation

    International Nuclear Information System (INIS)

    T. Lahnalampi; J. Case

    2005-01-01

    The purpose of the ''Closure and Sealing Design Calculation'' is to illustrate closure and sealing methods for sealing shafts, ramps, and identify boreholes that require sealing in order to limit the potential of water infiltration. In addition, this calculation will provide a description of the magma that can reduce the consequences of an igneous event intersecting the repository. This calculation will also include a listing of the project requirements related to closure and sealing. The scope of this calculation is to: summarize applicable project requirements and codes relating to backfilling nonemplacement openings, removal of uncommitted materials from the subsurface, installation of drip shields, and erecting monuments; compile an inventory of boreholes that are found in the area of the subsurface repository; describe the magma bulkhead feature and location; and include figures for the proposed shaft and ramp seals. The objective of this calculation is to: categorize the boreholes for sealing by depth and proximity to the subsurface repository; develop drawing figures which show the location and geometry for the magma bulkhead; include the shaft seal figures and a proposed construction sequence; and include the ramp seal figure and a proposed construction sequence. The intent of this closure and sealing calculation is to support the License Application by providing a description of the closure and sealing methods for the Safety Analysis Report. The closure and sealing calculation will also provide input for Post Closure Activities by describing the location of the magma bulkhead. This calculation is limited to describing the final configuration of the sealing and backfill systems for the underground area. The methods and procedures used to place the backfill and remove uncommitted materials (such as concrete) from the repository and detailed design of the magma bulkhead will be the subject of separate analyses or calculations. Post-closure monitoring will not

  9. The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

  10. Objectives of safety evaluation

    International Nuclear Information System (INIS)

    Rosen, M.

    1980-01-01

    An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants. In view of this and the generally small regulatory staffs of importing coutnries, suggestions are given for measures which should be taken by various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. (orig.)

  11. How safe is the safety paradigm?

    NARCIS (Netherlands)

    Arah, O. A.; Klazinga, N. S.

    2004-01-01

    This paper reviews safety initiatives in the health systems of the UK, Canada, Australia, and the US. Initiatives to tackle safety shortcomings involve public-private collaborations. Patient safety agencies (to institute learning, action and safety culture), adverse event reporting and, to a lesser

  12. Safety factors for neutron fluences in NPP safety assessment

    International Nuclear Information System (INIS)

    Demekhin, V.L.; Bukanov, V.N.; Il'kovich, V.V.; Pugach, A.M.

    2016-01-01

    In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25

  13. PROBABILISTIC MODEL FOR AIRPORT RUNWAY SAFETY AREAS

    Directory of Open Access Journals (Sweden)

    Stanislav SZABO

    2017-06-01

    Full Text Available The Laboratory of Aviation Safety and Security at CTU in Prague has recently started a project aimed at runway protection zones. The probability of exceeding by a certain distance from the runway in common incident/accident scenarios (take-off/landing overrun/veer-off, landing undershoot is being identified relative to the runway for any airport. As a result, the size and position of safety areas around runways are defined for the chosen probability. The basis for probability calculation is a probabilistic model using statistics from more than 1400 real-world cases where jet airplanes have been involved over the last few decades. Other scientific studies have contributed to understanding the issue and supported the model’s application to different conditions.

  14. Shielding calculations for NET

    International Nuclear Information System (INIS)

    Verschuur, K.A.; Hogenbirk, A.

    1991-05-01

    In the European Fusion Technology Programme there is only a small activity on research and development for fusion neutronics. Never-the-less, looking further than blanket design now, as ECN is getting involved in design of radiation shields for the coils and biological shields, it becomes apparent that fusion neutronics as a whole still needs substantial development. Existing exact codes for calculation of complex geometries like MCNP and DORT/TORT are put over the limits of their numerical capabilities, whilst approximate codes for complex geometries like FURNACE and MERCURE4 are put over the limits of their modelling capabilities. The main objective of this study is just to find out how far we can get with existing codes in obtaining reliable values for the radiation levels inside and outside the cryostat/shield during operation and after shut-down. Starting with a 1D torus model for preliminary parametric studies, more dimensional approximation of the torus or parts of it including the main heterogeneities should follow. Regular contacts with the NET-Team are kept, to be aware of main changes in NET design that might affect our calculation models. Work on the contract started 1 July 1990. The technical description of the contract is given. (author). 14 refs.; 4 figs.; 1 tab

  15. Accurate quantum chemical calculations

    Science.gov (United States)

    Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

    1989-01-01

    An important goal of quantum chemical calculations is to provide an understanding of chemical bonding and molecular electronic structure. A second goal, the prediction of energy differences to chemical accuracy, has been much harder to attain. First, the computational resources required to achieve such accuracy are very large, and second, it is not straightforward to demonstrate that an apparently accurate result, in terms of agreement with experiment, does not result from a cancellation of errors. Recent advances in electronic structure methodology, coupled with the power of vector supercomputers, have made it possible to solve a number of electronic structure problems exactly using the full configuration interaction (FCI) method within a subspace of the complete Hilbert space. These exact results can be used to benchmark approximate techniques that are applicable to a wider range of chemical and physical problems. The methodology of many-electron quantum chemistry is reviewed. Methods are considered in detail for performing FCI calculations. The application of FCI methods to several three-electron problems in molecular physics are discussed. A number of benchmark applications of FCI wave functions are described. Atomic basis sets and the development of improved methods for handling very large basis sets are discussed: these are then applied to a number of chemical and spectroscopic problems; to transition metals; and to problems involving potential energy surfaces. Although the experiences described give considerable grounds for optimism about the general ability to perform accurate calculations, there are several problems that have proved less tractable, at least with current computer resources, and these and possible solutions are discussed.

  16. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    computational tools and presentation of the results of the analysis. It also discusses various factors that need to be considered to ensure that the safety analysis is of an acceptable quality. In specific terms, the calculations and methods in this report can be used for the safety analysis of newly designed research reactors, modifications and experiments with impact on safety, and upgrades of existing reactors, and can also be used for updating or reassessing previous safety analyses of operating research reactors. This publication will be particularly useful to organizations, safety analysts and reviewers in fulfilling regulatory requirements and recommendations related to the preparation of the safety analysis and its presentation in the safety analysis report. In addition, it will help regulators conduct safety reviews and assessments of the topics covered

  17. Vaccine Safety

    Science.gov (United States)

    ... During Pregnancy Frequently Asked Questions about Vaccine Recalls Historical Vaccine Safety Concerns FAQs about GBS and Menactra ... CISA Resources for Healthcare Professionals Evaluation Current Studies Historical Background 2001-12 Publications Technical Reports Vaccine Safety ...

  18. SAFETY FIRST

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Ensuring safety while peacefully utilizing nuclear energy is a top priority for China A fter a recent earthquake in Japan caused radioactive leaks at a nuclear power plant in Tokyo, the safety of nuclear energy has again aroused public attention.

  19. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  20. Water Safety

    Science.gov (United States)

    ... Staying Safe Videos for Educators Search English Español Water Safety KidsHealth / For Parents / Water Safety What's in ... remains your best measure of protection. Making Kids Water Wise It's important to teach your kids proper ...

  1. Nuclear power and safety

    International Nuclear Information System (INIS)

    Chidambaram, R.

    1992-01-01

    Some aspects of safety of nuclear power with special reference to Indian nuclear power programme are discussed. India must develop technology to protect herself from the adverse economic impact arising out of the restrictive regime which is being created through globalization of safety and environmental issues. Though the studies done and experience gained so far have shown that the PHWR system adopted by India has a number of superior safety features, research work is needed in the field of operation and maintenance of reactors and also in the field of reactor life extension through delaying of ageing effects. Public relations work must be pursued to convince the public at large of the safety of nuclear power programme. The new reactor designs in the second stage of evolution are based on either further improvement of existing well-proven designs or adoptions of more innovative ideas based on physical principles to ensure a higher level of safety. The development of Indian nuclear power programme is characterised by a balanced approach in the matter of assuring safety. Safety enforcement is not just looked upon as a pure administrative matter, but experts with independent minds are also involved in safety related matters. (M.G.B.)

  2. Towards confidence in transport safety

    International Nuclear Information System (INIS)

    Robison, R.W.

    1992-01-01

    The U.S. Department of Energy (US DOE) plans to demonstrate to the public that high-level waste can be transported safely to the proposed repository. The author argues US DOE should begin now to demonstrate its commitment to safety by developing an extraordinary safety program for nuclear cargo it is now shipping. The program for current shipments should be developed with State, Tribal, and local officials. Social scientists should be involved in evaluating the effect of the safety program on public confidence. The safety program developed in cooperation with western states for shipments to the Waste Isolation Pilot plant is a good basis for designing that extraordinary safety program

  3. The International Criticality Safety Benchmark Evaluation Project

    International Nuclear Information System (INIS)

    Briggs, B. J.; Dean, V. F.; Pesic, M. P.

    2001-01-01

    In order to properly manage the risk of a nuclear criticality accident, it is important to establish the conditions for which such an accident becomes possible for any activity involving fissile material. Only when this information is known is it possible to establish the likelihood of actually achieving such conditions. It is therefore important that criticality safety analysts have confidence in the accuracy of their calculations. Confidence in analytical results can only be gained through comparison of those results with experimental data. The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the US Department of Energy. The project was managed through the Idaho National Engineering and Environmental Laboratory (INEEL), but involved nationally known criticality safety experts from Los Alamos National Laboratory, Lawrence Livermore National Laboratory, Savannah River Technology Center, Oak Ridge National Laboratory and the Y-12 Plant, Hanford, Argonne National Laboratory, and the Rocky Flats Plant. An International Criticality Safety Data Exchange component was added to the project during 1994 and the project became what is currently known as the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Representatives from the United Kingdom, France, Japan, the Russian Federation, Hungary, Kazakhstan, Korea, Slovenia, Yugoslavia, Spain, and Israel are now participating on the project In December of 1994, the ICSBEP became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency's (OECD-NEA) Nuclear Science Committee. The United States currently remains the lead country, providing most of the administrative support. The purpose of the ICSBEP is to: (1) identify and evaluate a comprehensive set of critical benchmark data; (2) verify the data, to the extent possible, by reviewing original and subsequently revised documentation, and by talking with the

  4. Food safety

    Science.gov (United States)

    ... safety URL of this page: //medlineplus.gov/ency/article/002434.htm Food safety To use the sharing features on this page, please enable JavaScript. Food safety refers to the conditions and practices that preserve the quality of food. These practices prevent contamination and foodborne ...

  5. Computing and physical methods to calculate Pu

    International Nuclear Information System (INIS)

    Mohamed, Ashraf Elsayed Mohamed

    2013-01-01

    Main limitations due to the enhancement of the plutonium content are related to the coolant void effect as the spectrum becomes faster, the neutron flux in the thermal region tends towards zero and is concentrated in the region from 10 Ke to 1 MeV. Thus, all captures by 240 Pu and 242 Pu in the thermal and epithermal resonance disappear and the 240 Pu and 242 Pu contributions to the void effect became positive. The higher the Pu content and the poorer the Pu quality, the larger the void effect. The core control in nominal or transient conditions Pu enrichment leads to a decrease in (B eff.), the efficiency of soluble boron and control rods. Also, the Doppler effect tends to decrease when Pu replaces U, so, that in case of transients the core could diverge again if the control is not effective enough. As for the voiding effect, the plutonium degradation and the 240 Pu and 242 Pu accumulation after multiple recycling lead to spectrum hardening and to a decrease in control. One solution would be to use enriched boron in soluble boron and shutdown rods. In this paper, I discuss and show the advanced computing and physical methods to calculate Pu inside the nuclear reactors and glovebox and the different solutions to be used to overcome the difficulties that effect, on safety parameters and on reactor performance, and analysis the consequences of plutonium management on the whole fuel cycle like Raw materials savings, fraction of nuclear electric power involved in the Pu management. All through two types of scenario, one involving a low fraction of the nuclear park dedicated to plutonium management, the other involving a dilution of the plutonium in all the nuclear park. (author)

  6. Magnetic Field Calculator

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Magnetic Field Calculator will calculate the total magnetic field, including components (declination, inclination, horizontal intensity, northerly intensity,...

  7. Smartphone apps for calculating insulin dose: a systematic assessment.

    Science.gov (United States)

    Huckvale, Kit; Adomaviciute, Samanta; Prieto, José Tomás; Leow, Melvin Khee-Shing; Car, Josip

    2015-05-06

    subtle harms resulting from suboptimal glucose control. Healthcare professionals should exercise substantial caution in recommending unregulated dose calculators to patients and address app safety as part of self-management education. The prevalence of errors attributable to incorrect interpretation of medical principles underlines the importance of clinical input during app design. Systemic issues affecting the safety and suitability of higher-risk apps may require coordinated surveillance and action at national and international levels involving regulators, health agencies and app stores.

  8. Safety handbook

    International Nuclear Information System (INIS)

    1990-01-01

    The purpose of the Australian Nuclear Science and Technology Organization's Safety Handbook is to outline simply the fundamental procedures and safety precautions which provide an appropriate framework for safe working with any potential hazards, such as fire and explosion, welding, cutting, brazing and soldering, compressed gases, cryogenic liquids, chemicals, ionizing radiations, non-ionising radiations, sound and vibration, as well as safety in the office. It also specifies the organisation for safety at the Lucas Heights Research Laboratories and the responsibilities of individuals and committees. It also defines the procedures for the scrutiny and review of all operations and the resultant setting of safety rules for them. ills

  9. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

    JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

  10. Fuel reprocessing: safety analysis of extraction cycles

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Nuclear Safety

    International Nuclear Information System (INIS)

    1978-09-01

    In this short paper it has only been possible to deal in a rather general way with the standards of safety used in the UK nuclear industry. The record of the industry extending over at least twenty years is impressive and, indeed, unique. No other industry has been so painstaking in protection of its workers and in its avoidance of damage to the environment. Headings are: introduction; how a nuclear power station works; radiation and its effects (including reference to ICRP, the UK National Radiological Protection Board, and safety standards); typical radiation doses (natural radiation, therapy, nuclear power programme and other sources); safety of nuclear reactors - design; key questions (matters of concern which arise in the public mind); safety of operators; safety of people in the vicinity of a nuclear power station; safety of the general public; safety bodies. (U.K.)

  12. Safety Behavior After Extinction Triggers a Return of Threat Expectancy

    NARCIS (Netherlands)

    van Uijen, S.L.; Leer, A.; Engelhard, I.M.

    2018-01-01

    Safety behavior is involved in the maintenance of anxiety disorders, presumably because it prevents the violation of negative expectancies. Recent research showed that safety behavior is resistant to fear extinction. This fear conditioning study investigated whether safety behavior after fear

  13. Criticality safety studies at VTT Energy

    International Nuclear Information System (INIS)

    Roine, T.; Anttila, M.

    1995-01-01

    At VTT Energy a compact reactor physics calculation system is applied in many kind of problems. Generation of group constants for static and dynamic core calculations, flux and dose rate calculations as well as criticality safety studies are performed basically with the same codes. In the presentation a short overview of the wide variety of criticality safety problems analyzed at VTT Energy is given. The calculation system with some illustrative examples is also described. (12 refs., 1 tab.)

  14. Fire safety

    International Nuclear Information System (INIS)

    Keski-Rahkonen, O.; Bjoerkman, J.; Hostikka, S.; Mangs, J.; Huhtanen, R.; Palmen, H.; Salminen, A.; Turtola, A.

    1998-01-01

    According to experience and probabilistic risk assessments, fires present a significant hazard in a nuclear power plant. Fires may be initial events for accidents or affect safety systems planned to prevent accidents and to mitigate their consequences. The project consists of theoretical work, experiments and simulations aiming to increase the fire safety at nuclear power plants. The project has four target areas: (1) to produce validated models for numerical simulation programmes, (2) to produce new information on the behavior of equipment in case of fire, (3) to study applicability of new active fire protecting systems in nuclear power plants, and (4) to obtain quantitative knowledge of ignitions induced by important electric devices in nuclear power plants. These topics have been solved mainly experimentally, but modelling at different level is used to interpret experimental data, and to allow easy generalisation and engineering use of the obtained data. Numerical fire simulation has concentrated in comparison of CFD modelling of room fires, and fire spreading on cables on experimental data. So far the success has been good to fair. A simple analytical and numerical model has been developed for fire effluents spreading beyond the room of origin in mechanically strongly ventilated compartments. For behaviour of equipment in fire several full scale and scaled down calorimetric experiments were carried out on electronic cabinets, as well as on horizontal and vertical cable trays. These were carried out to supply material for CFD numerical simulation code validation. Several analytical models were developed and validated against obtained experimental results to allow quick calculations for PSA estimates as well as inter- and extrapolations to slightly different objects. Response times of different commercial fire detectors were determined for different types of smoke, especially emanating from smoldering and flaming cables to facilitate selection of proper detector

  15. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  16. Safety climate and safety behaviors in the construction industry: The importance of co-workers commitment to safety.

    Science.gov (United States)

    Schwatka, Natalie V; Rosecrance, John C

    2016-06-16

    There is growing empirical evidence that as safety climate improves work site safety practice improve. Safety climate is often measured by asking workers about their perceptions of management commitment to safety. However, it is less common to include perceptions of their co-workers commitment to safety. While the involvement of management in safety is essential, working with co-workers who value and prioritize safety may be just as important. To evaluate a concept of safety climate that focuses on top management, supervisors and co-workers commitment to safety, which is relatively new and untested in the United States construction industry. Survey data was collected from a cohort of 300 unionized construction workers in the United States. The significance of direct and indirect (mediation) effects among safety climate and safety behavior factors were evaluated via structural equation modeling. Results indicated that safety climate was associated with safety behaviors on the job. More specifically, perceptions of co-workers commitment to safety was a mediator between both management commitment to safety climate factors and safety behaviors. These results support workplace health and safety interventions that build and sustain safety climate and a commitment to safety amongst work teams.

  17. Safety culture. Keys for sustaining progress

    International Nuclear Information System (INIS)

    Barraclough, I.; Carnino, A.

    1998-01-01

    Principles of nuclear safety are now well known and being put into practice around the world, leading to a degree of international harmonization in safety standards. Continued improvement in levels of safety requires the development of a comprehensive 'safety culture' at all levels of an organization, with visible and consistent leadership from senior management. This article reviews the main elements required for establishing and sustaining a good safety culture at nuclear installations that involves staff at all levels

  18. Organizational Culture and Safety

    Science.gov (United States)

    Adams, Catherine A.

    2003-01-01

    '..only a fool perseveres in error.' Cicero. Humans will break the most advanced technological devices and override safety and security systems if they are given the latitude. Within the workplace, the operator may be just one of several factors in causing accidents or making risky decisions. Other variables considered for their involvement in the negative and often catastrophic outcomes include the organizational context and culture. Many organizations have constructed and implemented safety programs to be assimilated into their culture to assure employee commitment and understanding of the importance of everyday safety. The purpose of this paper is to examine literature on organizational safety cultures and programs that attempt to combat vulnerability, risk taking behavior and decisions and identify the role of training in attempting to mitigate unsafe acts.

  19. Safety and environmental impacts

    International Nuclear Information System (INIS)

    Fiege, A.; Kramer, W.

    1991-01-01

    By means of interpreting experimental results, and by means of conservative estimates, several fundamental statements can be made concerning the safety and environmental impacts of fusion plants. Relevant findings so far regarding normal operation and incidents as well as risks involved in raw material extraction and waste management are compiled. (DG) [de

  20. CO2 flowrate calculator

    International Nuclear Information System (INIS)

    Carossi, Jean-Claude

    1969-02-01

    A CO 2 flowrate calculator has been designed for measuring and recording the gas flow in the loops of Pegase reactor. The analog calculator applies, at every moment, Bernoulli's formula to the values that characterize the carbon dioxide flow through a nozzle. The calculator electronics is described (it includes a sampling calculator and a two-variable function generator), with its amplifiers, triggers, interpolator, multiplier, etc. Calculator operation and setting are presented

  1. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

  2. Chemistry laboratory safety manual available

    Science.gov (United States)

    Elsbrock, R. G.

    1968-01-01

    Chemistry laboratory safety manual outlines safe practices for handling hazardous chemicals and chemistry laboratory equipment. Included are discussions of chemical hazards relating to fire, health, explosion, safety equipment and procedures for certain laboratory techniques and manipulations involving glassware, vacuum equipment, acids, bases, and volatile solvents.

  3. Radiation Safety Aspects of Nanotechnology

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, Mark [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, David; Cash, Leigh Jackson [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Guilmette, Raymond [Ray Guilmette & Associates, LLC, Perry, ME (United States); Kreyling, Wolfgang [Helmholtz-Zentrum Munchen, (Germany); Oberdorster, Gunter [Univ. of Rochester, NY (United States); Smith, Rachel [Public Health England, Oxfordshire (United Kingdom). Centre for Radiation, Chemical and Environmental Hazards

    2017-03-27

    This Report is intended primarily for operational health physicists, radiation safety officers, and internal dosimetrists who are responsible for establishing and implementing radiation safety programs involving radioactive nanomaterials. It should also provide useful information for workers, managers and regulators who are either working directly with or have other responsibilities related to work with radioactive nanomaterials.

  4. Safety culture

    International Nuclear Information System (INIS)

    1991-01-01

    The response to a previous publication by the International Nuclear Safety Advisory Group (INSAG), indicated a broad international interest in expansion of the concept of Safety Culture, in such a way that its effectiveness in particular cases may be judged. This report responds to that need. In its manifestation, Safety Culture has two major components: the framework determined by organizational policy and by managerial action, and the response of individuals in working within and benefiting by the framework. 1 fig

  5. Safety; Avertissement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  6. [Calculation of workers' health care costs].

    Science.gov (United States)

    Rydlewska-Liszkowska, Izabela

    2006-01-01

    In different health care systems, there are different schemes of organization and principles of financing activities aimed at ensuring the working population health and safety. Regardless of the scheme and the range of health care provided, economists strive for rationalization of costs (including their reduction). This applies to both employers who include workers' health care costs into indirect costs of the market product manufacture and health care institutions, which provide health care services. In practice, new methods of setting costs of workers' health care facilitate regular cost control, acquisition of detailed information about costs, and better adjustment of information to planning and control needs in individual health care institutions. For economic institutions and institutions specialized in workers' health care, a traditional cost-effect calculation focused on setting costs of individual products (services) is useful only if costs are relatively low and the output of simple products is not very high. But when products form aggregates of numerous actions like those involved in occupational medicine services, the method of activity based costing (ABC), representing the process approach, is much more useful. According to this approach costs are attributed to the product according to resources used during different activities involved in its production. The calculation of costs proceeds through allocation of all direct costs for specific processes in a given institution. Indirect costs are settled on the basis of resources used during the implementation of individual tasks involved in the process of making a new product. In this method, so called map of processes/actions consisted in the manufactured product and their interrelations are of particular importance. Advancements in the cost-effect for the management of health care institutions depend on their managerial needs. Current trends in this regard primarily depend on treating all cost reference

  7. Visit safety

    CERN Document Server

    2012-01-01

    Experiment areas, offices, workshops: it is possible to have co-workers or friends visit these places.     You already know about the official visits service, the VIP office, and professional visits. But do you know about the safety instruction GSI-OHS1, “Visits on the CERN site”? This is a mandatory General Safety Instruction that was created to assist you in ensuring safety for all your visits, whatever their nature—especially those that are non-official. Questions? The HSE Unit will be happy to answer them. Write to safety-general@cern.ch.   The HSE Unit

  8. Self-ignition of explosive substance. Comparison between analytical and numerical calculations in order to optimize safety in a pyrotechnic context; Auto-inflammation de substances explosives. Comparaison entre calcul analytique et numerique en vue d`une optimisation dans le domaine de la pyrotechnie

    Energy Technology Data Exchange (ETDEWEB)

    Gillard, Ph. [Centre National de la Recherche Scientifique (CNRS), 86 - Poitiers (France)

    1998-04-01

    Self-ignition of energetic material was investigated in order to optimize safety in the field of pyrotechnic applications. Two approaches were used; the first one is relative to Frank-Kamenetskii stationary thermal explosion theory. The second approach consists of a choice of some numerical solutions of heat conduction equations in a non-stationary state. Comparison between these results was carried out in order to find the numerical scheme which is the most compatible with Frank-Kamenetskii stationary thermal explosion theory. Numerical data were used for three explosive substances. One of them was studied by the author. In all cases, the numerical stationary state is in agreement with the Frank-Kamenetskii stationary thermal explosion theory, more or less accurately. From this comparison, it may be concluded that it is preferable, for this kind of problem, to use an implicit scheme with linearization of the heat source term. Explicit numerical methods, with or without the addition of the heat term with the Zinn and Mader scheme are revealed to be less accurate and to need a greater optimization of spatial and temporal meshing. (author) 7 refs.

  9. A meshless approach to radionuclide transport calculations

    International Nuclear Information System (INIS)

    Perko, J.; Sarler, B.

    2005-01-01

    Over the past thirty years numerical modelling has emerged as an interdisciplinary scientific discipline which has a significant impact in engineering and design. In the field of numerical modelling of transport phenomena in porous media, many commercial codes exist, based on different numerical methods. Some of them are widely used for performance assessment and safety analysis of radioactive waste repositories and groundwater modelling. Although they proved to be an accurate and reliable tool, they have certain limitations and drawbacks. Realistic problems often involve complex geometry which is difficult and time consuming to discretize. In recent years, meshless methods have attracted much attention due to their flexibility in solving engineering and scientific problems. In meshless methods the cumbersome polygonization of calculation domain is not necessary. By this the discretization time is reduced. In addition, the simulation is not as discretization density dependent as in traditional methods because of the lack of polygon interfaces. In this work fully meshless Diffuse Approximate Method (DAM) is used for calculation of radionuclide transport. Two cases are considered; First 1D comparison of 226 Ra transport and decay solved by the commercial Finite Volume Method (FVM) and Finite Element Method (FEM) based packages and DAM. This case shows the level of discretization density dependence. And second realistic 2D case of near-field modelling of radionuclide transport from the radioactive waste repository. Comparison is made again between FVM based code and DAM simulation for two radionuclides: Long-lived 14 C and short-lived 3 H. Comparisons indicate great capability of meshless methods to simulate complex transport problems and show that they should be seriously considered in future commercial simulation tools. (author)

  10. Operating experience: safety perspective

    International Nuclear Information System (INIS)

    Piplani, Vivek; Krishnamurthy, P.R.; Kumar, Neeraj; Upadhyay, Devendra

    2015-01-01

    Operating Experience (OE) provides valuable information for improving NPP safety. This may include events, precursors, deviations, deficiencies, problems, new insights to safety, good practices, lessons and corrective actions. As per INSAG-10, an OE program caters as a fundamental means for enhancing the defence-in-depth at NPPs and hence should be viewed as ‘Continuous Safety Performance Improvement Tool’. The ‘Convention on Nuclear Safety’ also recognizes the OE as a tool of high importance for enhancing the NPP safety and its Article 19 mandates each contracting party to establish an effective OE program at operating NPPs. The lessons drawn from major accidents at Three Mile Island, Chernobyl and Fukushima Daiichi NPPs had prompted nuclear stalwarts to change their safety perspective towards NPPs and to frame sound policies on issues like safety culture, severe accident prevention and mitigation. An effective OE program, besides correcting current/potential problems, help in proactively improving the NPP design, operating and maintenance procedures, practices, training, etc., and thus plays vital role in ensuring safe and efficient operation of NPPs. Further it enhances knowledge with regard to equipment operating characteristics, system performance trends and provides data for quantitative and qualitative safety analysis. Besides all above, an OE program inculcates a learning culture in the organisation and thus helps in continuously enhancing the expertise, technical competency and knowledge base of its staff. Nuclear and Radiation Facilities in India are regulated by Atomic Energy Regulatory Board (AERB). Operating Plants Safety Division (OPSD) of AERB is involved in managing operating experience activities. This paper provides insights about the operating experience program of OPSD, AERB (including its on-line data base namely OPSD STAR) and its utilisation in improving the regulations and safety at Indian NPPs/projects. (author)

  11. Assessment of Electrical Safety Beliefs and Practices: A Case Study

    Directory of Open Access Journals (Sweden)

    S. Boubaker

    2017-12-01

    Full Text Available In this paper, the electrical safety beliefs and practices in Hail region, Saudi Arabia, have been assessed. Based on legislative recommendations and rules applied in Saudi Arabia, on official statistics regarding the electricity-caused accidents and on the analysis of more than 200 photos captured in Hail (related to electrical safety, a questionnaire composed of 36 questions (10 for the respondents information, 16 for the home safety culture and 10 for the electrical devices purchasing culture has been devised and distributed to residents. 228 responses have been collected and analyzed. Using a scale similar to the one adopted for a university student GPA calculation, the electrical safety level (ESL in Hail region has been found to be 0.76 (in a scale of 4 points which is a very low score and indicates a poor electrical safety culture. Several recommendations involving different competent authorities have been proposed. Future work will concern the assessment of safety in industrial companies in Hail region.

  12. ICSBEP-2007, International Criticality Safety Benchmark Experiment Handbook

    International Nuclear Information System (INIS)

    Blair Briggs, J.

    2007-01-01

    1 - Description: The Critically Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United Sates Department of Energy. The project quickly became an international effort as scientist from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization of Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA). This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material. The example calculations presented do not constitute a validation of the codes or cross section data. The work of the ICSBEP is documented as an International Handbook of Evaluated Criticality Safety Benchmark Experiments. Currently, the handbook spans over 42,000 pages and contains 464 evaluations representing 4,092 critical, near-critical, or subcritical configurations and 21 criticality alarm placement/shielding configurations with multiple dose points for each and 46 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. The handbook is intended for use by criticality safety analysts to perform necessary validations of their calculational techniques and is expected to be a valuable tool for decades to come. The ICSBEP Handbook is available on DVD. You may request a DVD by completing the DVD Request Form on the internet. Access to the Handbook on the Internet requires a password. You may request a password by completing the Password Request Form. The Web address is: http://icsbep.inel.gov/handbook.shtml 2 - Method of solution: Experiments that are found

  13. Heterogeneous Calculation of {epsilon}

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, Alf

    1961-02-15

    A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of {epsilon}. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer.

  14. Heterogeneous Calculation of ε

    International Nuclear Information System (INIS)

    Jonsson, Alf

    1961-02-01

    A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of ε. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer

  15. Safety Principles

    Directory of Open Access Journals (Sweden)

    V. A. Grinenko

    2011-06-01

    Full Text Available The offered material in the article is picked up so that the reader could have a complete representation about concept “safety”, intrinsic characteristics and formalization possibilities. Principles and possible strategy of safety are considered. A material of the article is destined for the experts who are taking up the problems of safety.

  16. Safety Systems

    Science.gov (United States)

    Halligan, Tom

    2009-01-01

    Colleges across the country are rising to the task by implementing safety programs, response strategies, and technologies intended to create a secure environment for teachers and students. Whether it is preparing and responding to a natural disaster, health emergency, or act of violence, more schools are making campus safety a top priority. At…

  17. Safety First

    Science.gov (United States)

    Taft, Darryl

    2011-01-01

    Ned Miller does not take security lightly. As director of campus safety and emergency management at the Des Moines Area Community College (DMACC), any threat requires serious consideration. As community college administrators adopt a more proactive approach to campus safety, many institutions are experimenting with emerging technologies, including…

  18. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

  19. 46 CFR 174.360 - Calculations.

    Science.gov (United States)

    2010-10-01

    ... GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SUBDIVISION AND STABILITY SPECIAL RULES PERTAINING TO SPECIFIC VESSEL TYPES Special Rules Pertaining to Dry Cargo Ships § 174.360 Calculations. Each ship to... for that ship by the International Convention for the Safety of Life at Sea, 1974, as amended, chapter...

  20. 76 FR 71431 - Civil Penalty Calculation Methodology

    Science.gov (United States)

    2011-11-17

    ... DEPARTMENT OF TRANSPORTATION Federal Motor Carrier Safety Administration Civil Penalty Calculation... is currently evaluating its civil penalty methodology. Part of this evaluation includes a forthcoming... civil penalties. UFA takes into account the statutory penalty factors under 49 U.S.C. 521(b)(2)(D). The...

  1. Criticality criteria for submissions based on calculations

    International Nuclear Information System (INIS)

    Burgess, M.H.

    1975-06-01

    Calculations used in criticality clearances are subject to errors from various sources, and allowance must be made for these errors is assessing the safety of a system. A simple set of guidelines is defined, drawing attention to each source of error, and recommendations as to its application are made. (author)

  2. Fire safety engineering

    International Nuclear Information System (INIS)

    Smith, D.N.

    1989-01-01

    The periodic occurrence of large-scale, potentially disastrous industrial accidents involving fire in hazardous environments such as oilwell blowouts, petrochemical explosions and nuclear installations highlights the need for an integrated approach to fire safety engineering. Risk reduction 'by design' and rapid response are of equal importance in the saving of life and property in such situations. This volume of papers covers the subject thoroughly, touching on such topics as hazard analysis, safety design and testing, fire detection and control, and includes studies of fire hazard in the context of environment protection. (author)

  3. 1980 Annual status report reactor safety

    International Nuclear Information System (INIS)

    1981-01-01

    The JRC reactor safety programme involves theoretical and experimental activities to analyse accidents and their consequences for LWRs and LMFBRs. The first project deals with the improvement and the application of methodologies for risk and reliability assessment. This activity involves the identification and modelling of accident sequences and events and the analysis of fault trees. In this project, the implementation of a centralized data bank system (European Reliability Data System) is foreseen, which should provide the information needed for risk assessment studies. In project 2 a major effort on LWRs is centered on the study of the loss-of-coolant accident following large, intermediate or small breaks of the primary circuit. These accidents are simulated out of pile in the LOBI facility. In project 3 a contribution is made to solve material problems and to provide data and calculation methods for end of life predictions of reactor components. It involves a contribution to the programme for the inspection of steel components (PISC) as well as the study of fracture and creep fatigue properties of stainless steel. In the project 4 and 5 a deterministic approach is adopted to solve some problems of large hypothetical accidents in an LMFBR. The calculation tools developed concern sodium thermohydraulics in fuel element bundles, fuel coolant interaction, whole core accident analysis, containment loading and response and post accident heat removal

  4. Sensitivity analysis of reactor safety parameters with automated adjoint function generation

    International Nuclear Information System (INIS)

    Kallfelz, J.M.; Horwedel, J.E.; Worley, B.A.

    1992-01-01

    A project at the Paul Scherrer Institute (PSI) involves the development of simulation models for the transient analysis of the reactors in Switzerland (STARS). This project, funded in part by the Swiss Federal Nuclear Safety Inspectorate, also involves the calculation and evaluation of certain transients for Swiss light water reactors (LWRs). For best-estimate analyses, a key element in quantifying reactor safety margins is uncertainty evaluation to determine the uncertainty in calculated integral values (responses) caused by modeling, calculational methodology, and input data (parameters). The work reported in this paper is a joint PSI/Oak Ridge National Laboratory (ORNL) application to a core transient analysis code of an ORNL software system for automated sensitivity analysis. The Gradient-Enhanced Software System (GRESS) is a software package that can in principle enhance any code so that it can calculate the sensitivity (derivative) to input parameters of any integral value (response) calculated in the original code. The studies reported are the first application of the GRESS capability to core neutronics and safety codes

  5. Industrial safety: its structuring and content

    International Nuclear Information System (INIS)

    Munoz, A.; Rodriguez, J.; Martinez-Val, J.M.

    1999-01-01

    Industrial development has led to an on-going increase in productivity, but the concept of safety has also become highly relevant. In this article, the authors address the structuring and content of industrial safety which involves laying down essential safety requirements, both in manufacturing and processes and in products. (Author)

  6. Fundamental safety principles. Safety fundamentals

    International Nuclear Information System (INIS)

    2006-01-01

    This publication states the fundamental safety objective and ten associated safety principles, and briefly describes their intent and purpose. The fundamental safety objective - to protect people and the environment from harmful effects of ionizing radiation - applies to all circumstances that give rise to radiation risks. The safety principles are applicable, as relevant, throughout the entire lifetime of all facilities and activities - existing and new - utilized for peaceful purposes, and to protective actions to reduce existing radiation risks. They provide the basis for requirements and measures for the protection of people and the environment against radiation risks and for the safety of facilities and activities that give rise to radiation risks, including, in particular, nuclear installations and uses of radiation and radioactive sources, the transport of radioactive material and the management of radioactive waste

  7. Fundamental safety principles. Safety fundamentals

    International Nuclear Information System (INIS)

    2007-01-01

    This publication states the fundamental safety objective and ten associated safety principles, and briefly describes their intent and purpose. The fundamental safety objective - to protect people and the environment from harmful effects of ionizing radiation - applies to all circumstances that give rise to radiation risks. The safety principles are applicable, as relevant, throughout the entire lifetime of all facilities and activities - existing and new - utilized for peaceful purposes, and to protective actions to reduce existing radiation risks. They provide the basis for requirements and measures for the protection of people and the environment against radiation risks and for the safety of facilities and activities that give rise to radiation risks, including, in particular, nuclear installations and uses of radiation and radioactive sources, the transport of radioactive material and the management of radioactive waste

  8. Dimensions of Safety Climate among Iranian Nurses.

    Science.gov (United States)

    Konjin, Z Naghavi; Shokoohi, Y; Zarei, F; Rahimzadeh, M; Sarsangi, V

    2015-10-01

    Workplace safety has been a concern of workers and managers for decades. Measuring safety climate is crucial in improving safety performance. It is also a method of benchmarking safety perception. To develop and validate a psychometrics scale for measuring nurses' safety climate. Literature review, subject matter experts and nurse's judgment were used in items developing. Content validity and reliability for new tool were tested by content validity index (CVI) and test-retest analysis, respectively. Exploratory factor analysis (EFA) with varimax rotation was used to improve the interpretation of latent factors. A 40-item scale in 6 factors was developed, which could explain 55% of the observed variance. The 6 factors included employees' involvement in safety and management support, compliance with safety rules, safety training and accessibility to personal protective equipment, hindrance to safe work, safety communication and job pressure, and individual risk perception. The proposed scale can be used in identifying the needed areas to implement interventions in safety climate of nurses.

  9. The effect of an interactive e-drug calculations package on nursing students' drug calculation ability and self-efficacy.

    Science.gov (United States)

    McMullan, Miriam; Jones, Ray; Lea, Susan

    2011-06-01

    Nurses need to be competent and confident in performing drug calculations to ensure patient safety. The purpose of this study is to compare an interactive e-drug calculations package, developed using Cognitive Load Theory as its theoretical framework, with traditional handout learning support on nursing students' drug calculation ability, self-efficacy and support material satisfaction. A cluster randomised controlled trial comparing the e-package with traditional handout learning support was conducted with a September cohort (n=137) and a February cohort (n=92) of second year diploma nursing students. Students from each cohort were geographically dispersed over 3 or 4 independent sites. Students from each cohort were invited to participate, halfway through their second year, before and after a 12 week clinical practice placement. During their placement the intervention group received the e-drug calculations package while the control group received traditional 'handout' support material. Drug calculation ability and self-efficacy tests were given to the participants pre- and post-intervention. Participants were given the support material satisfaction scale post-intervention. Students in both cohorts randomised to e-learning were more able to perform drug calculations than those receiving the handout (September: mean 48.4% versus 34.7%, p=0.027; February: mean 47.6% versus 38.3%, p=0.024). February cohort students using the e-package were more confident in performing drug calculations than those students using handouts (self-efficacy mean 56.7% versus 45.8%, p=0.022). There was no difference in improved self-efficacy between intervention and control for students in the September cohort. Students who used the package were more satisfied with its use than the students who used the handout (mean 29.6 versus 26.5, p=0.001), particularly with regard to the package enhancing their learning (p=0.023), being an effective way to learn (p=0.005), providing practice and

  10. Safety strategy

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1980-01-01

    The basis for safety strategy in nuclear industry and especially nuclear power plants is the prevention of radioactivity release inside or outside of the technical installation. Therefore either technical or administrative measures are combined to a general strategy concept. This introduction will explain in more detail the following topics: - basic principles of safety - lines of assurance (LOA) - defense in depth - deterministic and probabilistic methods. This presentation is seen as an introduction to the more detailed discussion following in this course, nevertheless some selected examples will be used to illustrate the aspects of safety strategy development although they might be repeated later on. (orig.)

  11. Safety culture

    International Nuclear Information System (INIS)

    Drukraroff, C.

    2010-01-01

    The concept of Safety Culture was defined after Chernobyl's nuclear accident in 1986. It has not been exempt from discussion interpretations, adding riders, etc..., over the last 24 years because it has to do with human behavior and performance in the organizations. Safety Culture is not an easy task to define, assess and monitor. The proof of it is that today we still discussing and writing about it. How has been the evolution of Safety Culture at the Juzbado Factory since 1985 to today?. What is the strategy that we will be following in the future. (Author)

  12. Radiation safety

    International Nuclear Information System (INIS)

    1996-04-01

    Most of the ionizing radiation that people are exposed to in day-to-day activities comes from natural, rather than manmade, sources. The health effects of radiation - both natural and artificial - are relatively well understood and can be effectively minimized through careful safety measures and practices. The IAEA, together with other international and expert organizations, is helping to promote and institute Basic Safety Standards on an international basis to ensure that radiation sources and radioactive materials are managed for both maximum safety and human benefit

  13. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  14. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  15. HP-67 calculator programs for thermodynamic data and phase diagram calculations

    International Nuclear Information System (INIS)

    Brewer, L.

    1978-01-01

    This report is a supplement to a tabulation of the thermodynamic and phase data for the 100 binary systems of Mo with the elements from H to Lr. The calculations of thermodynamic data and phase equilibria were carried out from 5000 0 K to low temperatures. This report presents the methods of calculation used. The thermodynamics involved is rather straightforward and the reader is referred to any advanced thermodynamic text. The calculations were largely carried out using an HP-65 programmable calculator. In this report, those programs are reformulated for use with the HP-67 calculator; great reduction in the number of programs required to carry out the calculation results

  16. The International Technical Safety Forum

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    The International Technical Safety Forum is a meeting of safety experts from several physics labs in Europe and the US. Since 1998 participants have been meeting every couple of years to discuss common challenges in safety matters. The Forum helps them define best practices and learn from the important lessons learned by others.   The Forum's participants in front of building 40. This year, the meeting took place at CERN from 12 to 16 April. “This year's meeting covered subjects ranging from communication and training in matters of safety, to cryogenic safety, emergency preparedness and risk analysis”, explains Ralf Trant, head of the CERN Safety Commission and organiser of this year’s Forum. Radiation protection issues are not discussed at the meeting since they involve different expertise. The goal of the Forum is to allow participants to share experience, learn lessons and acquire specific knowledge in a very open way. Round-table discussions, dedicated time for ...

  17. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  18. Traffic safety strategies

    Directory of Open Access Journals (Sweden)

    V. Sadauskas

    2003-04-01

    Full Text Available Fast development of the number of vehicles is closely related not only to large benefit for the public but also to certain undesirable social and economic consequences. Firstly - large numbers of injured and killed people are involved into the accidents. The target to improve traffic safety situation in Lithuania can be reached only after the detailed evaluation of transport system, environment, traffic participants, road and vehicle. Taking into consideration the accident situation in Lithuania and its causes the followings priority trends are suggested: The improvement of the coordination of road traffic safety system, the training and education of road users, the explanation of the importance of traffic safety and its propagation, the improvement of traffic conditions. Recommendations and proposals for differentiated criterion of maximum speed limit selection taking into account different factors are provided in the work.

  19. International Criticality Safety Benchmark Evaluation Project (ICSBEP) - ICSBEP 2015 Handbook

    International Nuclear Information System (INIS)

    Bess, John D.

    2015-01-01

    The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy (DOE). The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirements and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross-section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span approximately 69000 pages and contain 567 evaluations with benchmark specifications for 4874 critical, near-critical or subcritical configurations, 31 criticality alarm placement/shielding configurations with multiple dose points for each, and 207 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the handbook are benchmark specifications for neutron activation foil and thermoluminescent dosimeter measurements performed at the SILENE critical assembly in Valduc, France as part of a joint venture in 2010 between the US DOE and the French Alternative Energies and Atomic Energy Commission (CEA). A photograph of this experiment is shown on the front cover. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these

  20. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  1. Further development and data basis for safety and accident analyses of nuclear front end and back end facilities and actualization and revision of calculation methods for nuclear safety analyses. Final report; Weiterentwicklung von Methoden und Datengrundlagen zu Sicherheits- und Stoerfallanalysen fuer Anlagen der nuklearen Ver- und Entsorgung sowie Aktualisierung und Ueberpruefung von Rechenmethoden zu nuklearen Sicherheitsanalysen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kilger, Robert; Peters, Elisabeth; Sommer, Fabian; Moser, Eberhard-Franz; Kessen, Sven; Stuke, Maik

    2016-07-15

    This report briefly describes the activities carried out under the project 3613R03350 on the GRS ''Handbook on Accident Analysis for Nuclear Front and Back End Facilities'', and in detail the continuing work on the revision and updating of the GRS ''Handbook on Criticality'', which here focused on fissile systems with plutonium and {sup 233}U. The in previous projects started and ongoing literature study on innovative fuel concepts is continued. Also described are the review and qualification of computational methods by research and active benchmark participation, and the results of tracking the state of science and technology in the field of computational methods for criticality safety analysis. Special in-depth analyzes of selected criticality-relevant occurrences in the past are also documented.

  2. Safety first!

    CERN Multimedia

    2016-01-01

    Among the many duties I assumed at the beginning of the year was the ultimate responsibility for Safety at CERN: the responsibility for the physical safety of the personnel, the responsibility for the safe operation of the facilities, and the responsibility to ensure that CERN acts in accordance with the highest standards of radiation and environmental protection.   The Safety Policy document drawn up in September 2014 is an excellent basis for the implementation of Safety in all areas of CERN’s work. I am happy to commit during my mandate to help meet its objectives, not least by ensuring the Organization makes available the necessary means to achieve its Safety objectives. One of the main objectives of the HSE (Occupational Health and Safety and Environmental Protection) unit in the coming months is to enhance the measures to minimise CERN’s impact on the environment. I believe CERN should become a role model for an environmentally-aware scientific research laboratory. Risk ...

  3. Electronics Environmental Benefits Calculator

    Data.gov (United States)

    U.S. Environmental Protection Agency — The Electronics Environmental Benefits Calculator (EEBC) was developed to assist organizations in estimating the environmental benefits of greening their purchase,...

  4. Electrical installation calculations basic

    CERN Document Server

    Kitcher, Christopher

    2013-01-01

    All the essential calculations required for basic electrical installation workThe Electrical Installation Calculations series has proved an invaluable reference for over forty years, for both apprentices and professional electrical installation engineers alike. The book provides a step-by-step guide to the successful application of electrical installation calculations required in day-to-day electrical engineering practice. A step-by-step guide to everyday calculations used on the job An essential aid to the City & Guilds certificates at Levels 2 and 3Fo

  5. Electrical installation calculations advanced

    CERN Document Server

    Kitcher, Christopher

    2013-01-01

    All the essential calculations required for advanced electrical installation workThe Electrical Installation Calculations series has proved an invaluable reference for over forty years, for both apprentices and professional electrical installation engineers alike. The book provides a step-by-step guide to the successful application of electrical installation calculations required in day-to-day electrical engineering practiceA step-by-step guide to everyday calculations used on the job An essential aid to the City & Guilds certificates at Levels 2 and 3For apprentices and electrical installatio

  6. Radar Signature Calculation Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: The calculation, analysis, and visualization of the spatially extended radar signatures of complex objects such as ships in a sea multipath environment and...

  7. Waste Package Lifting Calculation

    International Nuclear Information System (INIS)

    H. Marr

    2000-01-01

    The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation

  8. Laboratory safety handbook

    Science.gov (United States)

    Skinner, E.L.; Watterson, C.A.; Chemerys, J.C.

    1983-01-01

    Safety, defined as 'freedom from danger, risk, or injury,' is difficult to achieve in a laboratory environment. Inherent dangers, associated with water analysis and research laboratories where hazardous samples, materials, and equipment are used, must be minimized to protect workers, buildings, and equipment. Managers, supervisors, analysts, and laboratory support personnel each have specific responsibilities to reduce hazards by maintaining a safe work environment. General rules of conduct and safety practices that involve personal protection, laboratory practices, chemical handling, compressed gases handling, use of equipment, and overall security must be practiced by everyone at all levels. Routine and extensive inspections of all laboratories must be made regularly by qualified people. Personnel should be trained thoroughly and repetitively. Special hazards that may involve exposure to carcinogens, cryogenics, or radiation must be given special attention, and specific rules and operational procedures must be established to deal with them. Safety data, reference materials, and texts must be kept available if prudent safety is to be practiced and accidents prevented or minimized.

  9. Space station pressurized laboratory safety guidelines

    Science.gov (United States)

    Mcgonigal, Les

    1990-01-01

    Before technical safety guidelines and requirements are established, a common understanding of their origin and importance must be shared between Space Station Program Management, the User Community, and the Safety organizations involved. Safety guidelines and requirements are driven by the nature of the experiments, and the degree of crew interaction. Hazard identification; development of technical safety requirements; operating procedures and constraints; provision of training and education; conduct of reviews and evaluations; and emergency preplanning are briefly discussed.

  10. The evolution of cryogenic safety at Fermilab

    International Nuclear Information System (INIS)

    Stanek, R.; Kilmer, J.

    1992-12-01

    Over the past twenty-five years, Fermilab has been involved in cryogenic technology as it relates to pursuing experimentation in high energy physics. The Laboratory has instituted a strong cryogenic safety program and has maintained a very positive safety record. The solid commitment of management and the cryogenic community to incorporating safety into the system life cycle has led to policies that set requirements and help establish consistency for the purchase and installation of equipment and the safety analysis and documentation

  11. Supporting calculations and assumptions for use in WESF safetyanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hey, B.E.

    1997-03-07

    This document provides a single location for calculations and assumptions used in support of Waste Encapsulation and Storage Facility (WESF) safety analyses. It also provides the technical details and bases necessary to justify the contained results.

  12. Educational audit on drug dose calculation learning in a Tanzanian ...

    African Journals Online (AJOL)

    Background: Patient safety is a key concern for nurses; ability to calculate drug ... Specific objectives were to assess learning from targeted teaching, to identify problem areas in perfor- .... this could result in reduced risk of drug dose error in.

  13. Radiological safety and control

    International Nuclear Information System (INIS)

    Kim, Jang Hee; Kim, Ki Sub

    1995-01-01

    The practical objective of radiological safety control is intended for achievement and maintenance of appropreately safe condition in environmental control for activities involving exposure from the use of radiation. In order to establish these objectives, we should be to prevent deterministic effects and to limit the occurrence stochastic effects to level deemed to be acceptable by the application of general principles of radiation protection and systems of dose limitation based on ICRP recommendations. 34 tabs., 19 figs., 11 refs. (Author) .new

  14. Radiation safety and control

    International Nuclear Information System (INIS)

    Kim, Jang Hee; Kim, Gi Sub.

    1996-12-01

    The principal objective of radiological safety control is intended for achievement and maintenance of appropriately safe condition in environmental control for activities involving exposure from the use of radiation. In order to establish these objective, we should be to prevent deterministic effects and to limit the occurrence stochastic effects to level deemed to be acceptable by the application of general principles of radiation protection and systems of dose limitation based on ICRP recommendations. (author). 22 tabs., 13 figs., 11 refs

  15. Containment safety margins

    International Nuclear Information System (INIS)

    Von Riesemann, W.A.

    1980-01-01

    Objective of the Containment Safety Margins program is the development and verification of methodologies which are capable of reliably predicting the ultimate load-carrying capability of light water reactor containment structures under accident and severe environments. The program was initiated in June 1980 at Sandia and this paper addresses the first phase of the program which is essentially a planning effort. Brief comments are made about the second phase, which will involve testing of containment models

  16. Worker and public safety

    International Nuclear Information System (INIS)

    Hamel, P.E.

    1984-09-01

    Nuclear regulatory controls have been in place for many years in Canada to ensure that the risk for the safety of workers and members of the public is as low as reasonably possible. The Atomic Energy Control Board implements these controls by virtue of a broadly based Act of Parliament, rigorous regulations and compliance procedures. The Canadian experience with nuclear practices involves about 1 million person-years at risk without a fatality due to acute exposure to radiation

  17. Who and What Does Involvement Involve?

    DEFF Research Database (Denmark)

    Hansen, Jeppe Oute; Petersen, Anders; Huniche, Lotte

    2015-01-01

    This article gives an account of aspects of a multi-sited field study of involvement of relatives in Danish psychiatry. By following metaphors of involvement across three sites of the psychiatric systema family site, a clinical site and a policy sitethe first author (J.O.) investigated how...... theoretical perspective laid out by Ernesto Laclau and Chantal Mouffe, the aim of this study is to show how the dominant discourse about involvement at the political and clinical sites is constituted by understandings of mentally ill individuals and by political objectives of involvement. The analysis...... the responsibility toward the mental health of the ill individual as well as toward the psychological milieu of the family....

  18. Waste management safety

    International Nuclear Information System (INIS)

    Boehm, H.

    1983-01-01

    All studies carried out by competent authors of the safety of a waste management concept on the basis of reprocessing of the spent fuel elements and storage in the deep underground of the radioactive waste show that only a minor technical risk is involved in this step. This also holds true when evaluating the accidents which have occurred in waste management facilities. To explain the risk, first the completely different safety aspects of nuclear power plants, reprocessing plants and repositories are outlined together with the safety related characteristics of these plants. Also this comparison indicates that the risk of waste management facilities is considerably lower than the, already very small, risk of nuclear power plants. For the final storage of waste from reprocessing and for the direct storage of fuel elements, the results of safety analyses show that the radiological exposure following an accident with radioactivity releases, even under conservative assumptions, is considerably below the natural radiation exposure. The very small danger to the environment arising from waste management by reprocessing clearly indicates that aspects of technical safety alone will hardly be a major criterion for the decision in favor of one or the other waste management approach. (orig.) [de

  19. Ignalina Safety Analysis Group

    International Nuclear Information System (INIS)

    Ushpuras, E.

    1995-01-01

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

  20. 3D analysis methods - Study and seminar[BWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Daaviittila, A [Valtion Teknillinen Tutkimuskeskus (Finland)

    2003-10-01

    The first part of the report results from a study that was performed as a Nordic co-operation activity with active participation from Studsvik Scandpower and Westinghouse Atom in Sweden, and VTT in Finland. The purpose of the study was to identify and investigate the effects rising from using the 3D transient com-puter codes in BWR safety analysis, and their influence on the transient analysis methodology. One of the main questions involves the critical power ratio (CPR) calculation methodology. The present way, where the CPR calculation is per-formed with a separate hot channel calculation, can be artificially conservative. In the investigated cases, no dramatic minimum CPR effect coming from the 3D calculation is apparent. Some cases show some decrease in the transient change of minimum CPR with the 3D calculation, which confirms the general thinking that the 1D calculation is conservative. On the other hand, the observed effect on neutron flux behaviour is quite large. In a slower transient the 3D effect might be stronger. The second part of the report is a summary of a related seminar that was held on the 3D analysis methods. The seminar was sponsored by the Reactor Safety part (NKS-R) of the Nordic Nuclear Safety Research Programme (NKS). (au)

  1. Initialization of Safety Assessment Process for the Croatian Radioactive Waste repository on Trgovska gora

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.; Subasic, D.

    2000-01-01

    An iterative process of safety assessment, presently focusing on the site-specific evaluation of the post-closure phase for the prospective LILW repository on Trgovska gora in Croatia, has recently been initiated. The primary aim of the first assessment iterations is to provide the experts involved, the regulators and the general public with a reasonable assurance that the applicable long term performance and safety objectives can be met. Another goal is to develop a sufficient understanding of the system behavior to support decisions about the site investigation, the facility design, the waste acceptance criteria and the closure conditions. In this initial phase, the safety assessment is structured in a manner following closely methodology of the ISAM. The International Programme for Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities the IAEA coordinated research program started in 1997. Results of the safety assessment first iteration will be organized and presented in the form of a preliminary safety analysis report (PSAR), expected to be completed in the second part of the year 2000. As the first report on the initiated safety assessment activities, the PSAR will describe the concept and aims of the assessment process. Particular emphasis will be placed on description of the key elements of a safety assessment approach by: a) defining the assessment context; b) providing description of the disposal system; c) developing and justifying assessment scenarios; d) formulating and implementing models; and e) interpreting the scoping calculations. (author)

  2. Whole core burnup calculations using 'MCNP'

    International Nuclear Information System (INIS)

    Haran, O.; Shaham, Y.

    1996-01-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

  3. Whole core burnup calculations using `MCNP`

    Energy Technology Data Exchange (ETDEWEB)

    Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

  4. Mechanical calculation of heat exchangers

    International Nuclear Information System (INIS)

    Osweiller, Francis.

    1977-01-01

    Many heat exchangers are still being dimensioned at the present time by means of the American TEMA code (Tubular Exchanger Manufacturers Association). The basic formula of this code often gives rise to significant tubular plate thicknesses which, apart from the cost of materials, involve significant machining. Some constructors have brought into use calculation methods that are more analytic so as to take into better consideration the mechanical phenomena which come into play in a heat exchanger. After a brief analysis of these methods it is shown, how the original TEMA formulations have changed to reach the present version and how this code has incorporated Gardner's results for treating exchangers with two fixed heads. A formal and numerical comparison is then made of the analytical and TEMA methods by attempting to highlight a code based on these methods or a computer calculation programme in relation to the TEMA code [fr

  5. SAFETY INSTRUCTION AND SAFETY NOTE

    CERN Multimedia

    TIS Secretariat

    2002-01-01

    Please note that the SAFETY INSTRUCTION N0 49 (IS 49) and the SAFETY NOTE N0 28 (NS 28) entitled respectively 'AVOIDING CHEMICAL POLLUTION OF WATER' and 'CERN EXHIBITIONS - FIRE PRECAUTIONS' are available on the web at the following urls: http://edms.cern.ch/document/335814 and http://edms.cern.ch/document/335861 Paper copies can also be obtained from the TIS Divisional Secretariat, email: TIS.Secretariat@cern.ch

  6. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  7. Uneconomical top calculation method

    International Nuclear Information System (INIS)

    De Noord, M.; Vanm Sambeek, E.J.W.

    2003-08-01

    The methodology used to calculate the financial gap of renewable electricity sources and technologies is described. This methodology is used for calculating the production subsidy levels (MEP subsidies) for new renewable electricity projects in 2004 and 2005 in the Netherlands [nl

  8. Calculation of dietary exposure to acrylamide in the Norwegian population

    OpenAIRE

    Norwegian Scientific Committee for Food Safety

    2015-01-01

    The Norwegian Scientific Committee for Food Safety (VKM) is requested by the Norwegian Food Safety Authority (NFSA) to calculate the dietary exposure to acrylamide in the Norwegian population. NFSA refers to the recent scientific opinion on acrylamide in food by the European Food Safety Authority (EFSA). EFSA concludes that acrylamide in food potentially increases the risk of developing cancer for consumers in all age groups.

  9. 47 CFR 36.605 - Calculation of safety net additive.

    Science.gov (United States)

    2010-10-01

    ..., REVENUES, EXPENSES, TAXES AND RESERVES FOR TELECOMMUNICATIONS COMPANIES 1 Universal Service Fund General..., the rural incumbent local exchange carrier realizes growth in end of period Telecommunications Plant... percent more than the study area's TPIS per loop investment at the end of the prior period. (2) If...

  10. Nomograms for calculating the safety factor of homogeneous earth ...

    African Journals Online (AJOL)

    use

    according the material classification and the parameters of design, height and slope. Key words: Nomograms ... works, mechanical properties of materials, and software employed. (Lakehal, 2008). .... Engineering Manual. Engineering and ...

  11. Safety concerns with the Centers for Disease Control opioid calculator

    Directory of Open Access Journals (Sweden)

    Fudin J

    2017-12-01

    Full Text Available Jeffrey Fudin,1–4 Mena Raouf,2 Erica L Wegrzyn,2–4 Michael E Schatman5,61Scientific and Clinical Affairs, Remitigate, LLC, Delmar, NY, USA; 2Stratton VA Medical Center, Albany, NY, USA; 3Western New England University College of Pharmacy, Springfield, MA, USA; 4Albany College of Pharmacy & Health Sciences, Albany, NY, USA; 5Research and Network Development, Boston Pain Care, Waltham, MA, USA; 6Department of Public Health & Community Medicine, Tufts University School of Medicine, Boston, MA, USAMorphine milligram equivalence (MME and other comparable acronyms have been employed in federal pain guidelines and used by policy makers to limit opioid prescribing.1–5 On March 18, 2016, the Centers for Disease Control (CDC released its Guideline for Prescribing Opioids for Chronic Pain.1 The guidelines provided 12 recommendations for “primary care clinicians prescribing opioids for chronic pain outside of active cancer treatment, palliative care, and end-of-life care”. One of the CDC recommendations states that clinicians “should avoid increasing dosage to ≥90 MME/day or carefully justify a decision to titrate dosage to ≥90 MME/day”.1

  12. Nuclear safety-related calculations for Ghana Research Reactor -1 ...

    African Journals Online (AJOL)

    , kinet-ic parameters and isothermal reactivity coefficients. Comparisons of some computed values with results obtain-ed using other codes, which are in good agreement, are presented. Journal of Applied Science and Technology (JAST) , Vol.

  13. French concepts of ''passive safety''

    International Nuclear Information System (INIS)

    Dennielou, Y.; Serret, M.

    1990-01-01

    N 4 model, the French 1400 MW PWR of the 90's, exhibits many advanced features. As far as safety is concerned, the fully computerized control room design takes advantage of the operating experience feedback and largely improves the man machine interface. New post-accident procedures have been developed (the so-called ''physical states oriented procedures''). A complete consistent set of ''Fundamental Safety Rules'' have been issued. This however doesn't imply any significant modification of standard PWR with regard to the passive aspects of safety systems or functions. Nevertheless, traditional PWR safety systems largely use passive aspects: natural circulation, reactivity coefficients, gravity driven control rods, injection accumulators, so on. Moreover, probability calculations allow for comparison between the respective contributions of passive and of active failures. In the near future, eventual options of future French PWRs to be commissioned after 2000 will be evaluated; simplification, passive and forgiving aspects of safety systems will be thoroughly considered. (author)

  14. Application of the perturbation theory for sensitivity calculations in thermalhydraulics reactor calculations

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de

    1986-01-01

    The sensitivity of non linear responses associated with physical quantities governed by non linear differential systems can be studied using perturbation theory. The equivalence and formal differences between the differential and GPT formalisms are shown and both are used for sensitivity calculations of transient problems in a typical PWR coolant channel. The results obtained are encouraging with respect to the potential of the method for thermalhydraulics calculations normally performed for reactor design and safety analysis. (Author) [pt

  15. Methods for calculating nonconcave entropies

    International Nuclear Information System (INIS)

    Touchette, Hugo

    2010-01-01

    Five different methods which can be used to analytically calculate entropies that are nonconcave as functions of the energy in the thermodynamic limit are discussed and compared. The five methods are based on the following ideas and techniques: (i) microcanonical contraction, (ii) metastable branches of the free energy, (iii) generalized canonical ensembles with specific illustrations involving the so-called Gaussian and Betrag ensembles, (iv) the restricted canonical ensemble, and (v) the inverse Laplace transform. A simple long-range spin model having a nonconcave entropy is used to illustrate each method

  16. Dose calculation for electrons

    International Nuclear Information System (INIS)

    Hirayama, Hideo

    1995-01-01

    The joint working group of ICRP/ICRU is advancing the works of reviewing the ICRP publication 51 by investigating the data related to radiation protection. In order to introduce the 1990 recommendation, it has been demanded to carry out calculation for neutrons, photons and electrons. As for electrons, EURADOS WG4 (Numerical Dosimetry) rearranged the data to be calculated at the meeting held in PTB Braunschweig in June, 1992, and the question and request were presented by Dr. J.L. Chartier, the responsible person, to the researchers who are likely to undertake electron transport Monte Carlo calculation. The author also has carried out the requested calculation as it was the good chance to do the mutual comparison among various computation codes regarding electron transport calculation. The content that the WG requested to calculate was the absorbed dose at depth d mm when parallel electron beam enters at angle α into flat plate phantoms of PMMA, water and ICRU4-element tissue, which were placed in vacuum. The calculation was carried out by the versatile electron-photon shower computation Monte Carlo code, EGS4. As the results, depth dose curves and the dependence of absorbed dose on electron energy, incident angle and material are reported. The subjects to be investigated are pointed out. (K.I.)

  17. Handbook for the calculation of reactor protections; Formulaire sur le calcul de la protection des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-07-01

    This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

  18. Radiation safety

    International Nuclear Information System (INIS)

    Van Riessen, A.

    2002-01-01

    Full text: Experience has shown that modem, fully enclosed, XRF and XRD units are generally safe. This experience may lead to complacency and ultimately a lowering of standards which may lead to accidents. Maintaining awareness of radiation safety issues is thus an important role for all radiation safety officers. With the ongoing progress in technology, a greater number of radiation workers are more likely to use a range of instruments/techniques - eg portable XRF, neutron beam analysis, and synchrotron radiation analysis. The source for each of these types of analyses is different and necessitates an understanding of the associated dangers as well as use of specific radiation badges. The trend of 'suitcase science' is resulting in scientists receiving doses from a range of instruments and facilities with no coordinated approach to obtain an integrated dose reading for an individual. This aspect of radiation safety needs urgent attention. Within Australia a divide is springing up between those who work on Commonwealth property and those who work on State property. For example a university staff member may operate irradiating equipment on a University campus and then go to a CSIRO laboratory to operate similar equipment. While at the University State regulations apply and while at CSIRO Commonwealth regulations apply. Does this individual require two badges? Is there a need to obtain two licences? The application of two sets of regulations causes unnecessary confusion and increases the workload of radiation safety officers. Radiation safety officers need to introduce risk management strategies to ensure that both existing and new procedures result in risk minimisation. A component of this strategy includes ongoing education and revising of regulations. AXAA may choose to contribute to both of these activities as a service to its members as well as raising the level of radiation safety for all radiation workers. Copyright (2002) Australian X-ray Analytical

  19. The spin project: safety and performance indicators in different time frames

    International Nuclear Information System (INIS)

    Storck, R.; Becker, D.A.

    2002-01-01

    Safety and performance indicators have been under discussion for many years in several countries and international organisations. If those indicators refer to the long term safety of the total disposal system, they are often called safety indicators. If they refer to the performance of subsystems or the total system from a more technical point of view, they are sometimes called performance indicators. The need for indicators other than dose rates derives e.g. from the long time frames involved in safety assessments of waste disposal systems and the increasing uncertainty in dose rate calculations over time due to uncertainty in evolution of the surface environment and of behaviour of man. Before introducing additional indicators into a safety case of a potential repository site, the applicability and usefulness of different indicators have to be investigated and evaluated. The systematic analysis and testing of safety and performance indicators for use in different time horizons after closure of the disposal facility is the task of the SPIN project. This is done by re-calculating four recent studies concerning repository projects in granite formations. (authors)

  20. Large scale GW calculations

    International Nuclear Information System (INIS)

    Govoni, Marco; Argonne National Lab., Argonne, IL; Galli, Giulia; Argonne National Lab., Argonne, IL

    2015-01-01

    We present GW calculations of molecules, ordered and disordered solids and interfaces, which employ an efficient contour deformation technique for frequency integration and do not require the explicit evaluation of virtual electronic states nor the inversion of dielectric matrices. We also present a parallel implementation of the algorithm, which takes advantage of separable expressions of both the single particle Green's function and the screened Coulomb interaction. The method can be used starting from density functional theory calculations performed with semilocal or hybrid functionals. The newly developed technique was applied to GW calculations of systems of unprecedented size, including water/semiconductor interfaces with thousands of electrons

  1. Radioactive cloud dose calculations

    International Nuclear Information System (INIS)

    Healy, J.W.

    1984-01-01

    Radiological dosage principles, as well as methods for calculating external and internal dose rates, following dispersion and deposition of radioactive materials in the atmosphere are described. Emphasis has been placed on analytical solutions that are appropriate for hand calculations. In addition, the methods for calculating dose rates from ingestion are discussed. A brief description of several computer programs are included for information on radionuclides. There has been no attempt to be comprehensive, and only a sampling of programs has been selected to illustrate the variety available

  2. Safety organization

    International Nuclear Information System (INIS)

    Lutz, M.

    1984-06-01

    After a rapid definition of a nuclear basis installation, the national organization of nuclear safety in France is presented, as also the main organizations concerned and their functions. This report shows how the licensing procedure leading to the construction and exploitation of such installations is applied in the case of nuclear laboratories of research and development: examinations of nuclear safety problems are carried out at different levels: - centralized to define the frame out of which the installation has not to operate, - decentralized to follow in a more detailed manner its evolution [fr

  3. Operational safety

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The PNL Safety, Standards and Compliance Program contributed to the development and issuance of safety policies, standards, and criteria; for projects in the nuclear and nonnuclear areas. During 1976 the major emphasis was on developing criteria, instruments and methods to assure that radiation exposure to occupational personnel and to people in the environs of nuclear-related facilities is maintained at the lowest level technically and economically practicable. Progress in 1976 is reported on the preparation of guidelines for radiation exposure; Pu dosimetry studies; the preparation of an environmental monitoring handbook; and emergency instrumentation preparedness

  4. Neutronics calculation of RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

    2017-01-01

    Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

  5. Prenatal radiation exposure. Dose calculation

    International Nuclear Information System (INIS)

    Scharwaechter, C.; Schwartz, C.A.; Haage, P.; Roeser, A.

    2015-01-01

    The unborn child requires special protection. In this context, the indication for an X-ray examination is to be checked critically. If thereupon radiation of the lower abdomen including the uterus cannot be avoided, the examination should be postponed until the end of pregnancy or alternative examination techniques should be considered. Under certain circumstances, either accidental or in unavoidable cases after a thorough risk assessment, radiation exposure of the unborn may take place. In some of these cases an expert radiation hygiene consultation may be required. This consultation should comprise the expected risks for the unborn while not perturbing the mother or the involved medical staff. For the risk assessment in case of an in-utero X-ray exposition deterministic damages with a defined threshold dose are distinguished from stochastic damages without a definable threshold dose. The occurrence of deterministic damages depends on the dose and the developmental stage of the unborn at the time of radiation. To calculate the risks of an in-utero radiation exposure a three-stage concept is commonly applied. Depending on the amount of radiation, the radiation dose is either estimated, roughly calculated using standard tables or, in critical cases, accurately calculated based on the individual event. The complexity of the calculation thereby increases from stage to stage. An estimation based on stage one is easily feasible whereas calculations based on stages two and especially three are more complex and often necessitate execution by specialists. This article demonstrates in detail the risks for the unborn child pertaining to its developmental phase and explains the three-stage concept as an evaluation scheme. It should be noted, that all risk estimations are subject to considerable uncertainties.

  6. Calculating zeros: Non-equilibrium free energy calculations

    International Nuclear Information System (INIS)

    Oostenbrink, Chris; Gunsteren, Wilfred F. van

    2006-01-01

    Free energy calculations on three model processes with theoretically known free energy changes have been performed using short simulation times. A comparison between equilibrium (thermodynamic integration) and non-equilibrium (fast growth) methods has been made in order to assess the accuracy and precision of these methods. The three processes have been chosen to represent processes often observed in biomolecular free energy calculations. They involve a redistribution of charges, the creation and annihilation of neutral particles and conformational changes. At very short overall simulation times, the thermodynamic integration approach using discrete steps is most accurate. More importantly, reasonable accuracy can be obtained using this method which seems independent of the overall simulation time. In cases where slow conformational changes play a role, fast growth simulations might have an advantage over discrete thermodynamic integration where sufficient sampling needs to be obtained at every λ-point, but only if the initial conformations do properly represent an equilibrium ensemble. From these three test cases practical lessons can be learned that will be applicable to biomolecular free energy calculations

  7. Radiation Safety (Qualifications) Regulations 1980

    International Nuclear Information System (INIS)

    1980-01-01

    These Regulations, promulgated pursuant to the provisions of the Radiation Safety Act, 1975-1979, require persons engaged in activities involving radiation to pass a radiation safety examination or to possess an approved qualification in radiation. The National Health and Medical Research Council is authorised to exempt persons from compliance with these requirements or, conversely, to impose such requirements on persons other than those designated. (NEA) [fr

  8. Handout on shielding calculation

    International Nuclear Information System (INIS)

    Heilbron Filho, P.F.L.

    1991-01-01

    In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)

  9. Unit Cost Compendium Calculations

    Data.gov (United States)

    U.S. Environmental Protection Agency — The Unit Cost Compendium (UCC) Calculations raw data set was designed to provide for greater accuracy and consistency in the use of unit costs across the USEPA...

  10. PHYSICOCHEMICAL PROPERTY CALCULATIONS

    Science.gov (United States)

    Computer models have been developed to estimate a wide range of physical-chemical properties from molecular structure. The SPARC modeling system approaches calculations as site specific reactions (pKa, hydrolysis, hydration) and `whole molecule' properties (vapor pressure, boilin...

  11. Magnetic Field Grid Calculator

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Magnetic Field Properties Calculator will computes the estimated values of Earth's magnetic field(declination, inclination, vertical component, northerly...

  12. Intercavitary implants dosage calculation

    International Nuclear Information System (INIS)

    Rehder, B.P.

    The use of spacial geometry peculiar to each treatment for the attainment of intercavitary and intersticial implants dosage calculation is presented. The study is made in patients with intercavitary implants by applying a modified Manchester technique [pt

  13. Casio Graphical Calculator Project.

    Science.gov (United States)

    Stott, Nick

    2001-01-01

    Shares experiences of a project aimed at developing and refining programs written on a Casio FX9750G graphing calculator. Describes in detail some programs used to develop mental strategies and problem solving skills. (MM)

  14. Small portable speed calculator

    Science.gov (United States)

    Burch, J. L.; Billions, J. C.

    1973-01-01

    Calculator is adapted stopwatch calibrated for fast accurate measurement of speeds. Single assembled unit is rugged, self-contained, and relatively inexpensive to manufacture. Potential market includes automobile-speed enforcement, railroads, and field-test facilities.

  15. Calculativeness and trust

    DEFF Research Database (Denmark)

    Frederiksen, Morten

    2014-01-01

    Williamson’s characterisation of calculativeness as inimical to trust contradicts most sociological trust research. However, a similar argument is found within trust phenomenology. This paper re-investigates Williamson’s argument from the perspective of Løgstrup’s phenomenological theory of trust....... Contrary to Williamson, however, Løgstrup’s contention is that trust, not calculativeness, is the default attitude and only when suspicion is awoken does trust falter. The paper argues that while Williamson’s distinction between calculativeness and trust is supported by phenomenology, the analysis needs...... to take actual subjective experience into consideration. It points out that, first, Løgstrup places trust alongside calculativeness as a different mode of engaging in social interaction, rather conceiving of trust as a state or the outcome of a decision-making process. Secondly, the analysis must take...

  16. IRIS core criticality calculations

    International Nuclear Information System (INIS)

    Jecmenica, R.; Trontl, K.; Pevec, D.; Grgic, D.

    2003-01-01

    Three-dimensional Monte Carlo computer code KENO-VI of CSAS26 sequence of SCALE-4.4 code system was applied for pin-by-pin calculations of the effective multiplication factor for the first cycle IRIS reactor core. The effective multiplication factors obtained by the above mentioned Monte Carlo calculations using 27-group ENDF/B-IV library and 238-group ENDF/B-V library have been compared with the effective multiplication factors achieved by HELIOS/NESTLE, CASMO/SIMULATE, and modified CORD-2 nodal calculations. The results of Monte Carlo calculations are found to be in good agreement with the results obtained by the nodal codes. The discrepancies in effective multiplication factor are typically within 1%. (author)

  17. Relative Effects of Psychological Flexibility, Parental Involvement ...

    African Journals Online (AJOL)

    A critical analysis and understanding of secondary students' experiences and of safety in public schools are currently lacking in the literature and warrant further research. This study investigated the relative effects of psychological flexibility, parental involvement and school climate on secondary school student's school ...

  18. Safety control and risk management

    International Nuclear Information System (INIS)

    Rasmussen, J.

    1987-01-01

    The acceptable probability of major accidents in nuclear power is very small, and can not be determined from direct empirical evidence. Therefore, control of the level of safety is a complex problem. The difficulty is related to the fact that a variable, 'safety', which is not accessible to direct measurement, is to be tightly controlled. Control, therefore, depends on a systematic, analytical prediction of the target state, i.e., the level of safety, from indirect evidence. From a control theoretic point of view this means that safety is controlled by a system which includes openloop as well as closed loop control paths. The aim of the paper is to take a general systems view on the complex mechanisms involved in the control of safety of industrial installations like nuclear power. From this, the role of probabilistic risk analysis is evaluated and needs for further development discussed. (author)

  19. Source and replica calculations

    International Nuclear Information System (INIS)

    Whalen, P.P.

    1994-01-01

    The starting point of the Hiroshima-Nagasaki Dose Reevaluation Program is the energy and directional distributions of the prompt neutron and gamma-ray radiation emitted from the exploding bombs. A brief introduction to the neutron source calculations is presented. The development of our current understanding of the source problem is outlined. It is recommended that adjoint calculations be used to modify source spectra to resolve the neutron discrepancy problem

  20. Shielding calculations using FLUKA

    International Nuclear Information System (INIS)

    Yamaguchi, Chiri; Tesch, K.; Dinter, H.

    1988-06-01

    The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)

  1. Reactor safety research - results and perspectives

    International Nuclear Information System (INIS)

    Banaschik, M.

    1989-01-01

    The work performed so far is an essential contribution to the determination of the safety margins of nuclear facilities and their systems and to the further development of safety engineering. The further development of safety engineering involves a shift of emphasis in reactor safety research towards event sequences beyond the design basis. The aim of this shift in emphasis is the further development of the preventive level. This is based on the fact that the conservative design of the operating and safety systems involves and essential safety potential. The R and D work is intended to help develop accident management measures and to take the plant back into the safe state even after severe accidents. In this context, it is necessary to make full use of the safety margins of the plant and to include the operating systems for coping with accidents. As a result of the aims, the research work approaches operating and plant-specific processes. (orig./DG) [de

  2. General principles of nuclear safety management related to research reactor decommissioning

    International Nuclear Information System (INIS)

    Banciu, Ortenzia; Vladescu, Gabriela

    2003-01-01

    The paper contents the general principles applicable to the decommissioning of research reactors to ensure a proper nuclear safety management, during both decommissioning activities and post decommissioning period. The main objective of decommissioning is to ensure the protection of workers, population and environment against all radiological and non-radiological hazards that could result after a reactor shutdown and dismantling. In the same time, it is necessary, by some proper provisions, to limit the effect of decommissioning for the future generation, according to the new Romanian, IAEA and EU Norms and Regulations. Assurance of nuclear safety during decommissioning process involves, in the first step, to establish of some safety principles and requirements to be taken into account during whole process. In the same time, it is necessary to perform a series of analyses to ensure that the whole process is conducted in a planned and safe manner. The general principles proposed for a proper management of safety during research reactor decommissioning are as follows: - Set-up of all operations included in a Decommissioning Plan; - Set-up and qualitative evaluation of safety problems, which could appear during normal decommissioning process, both radiological and nonradiological risks for workers and public; - Set-up of accident list related to decommissioning process the events that could appear both due to some abnormal working conditions and to some on-site and off-site events like fires, explosions, flooding, earthquake, etc.); - Development and qualitative/ quantitative evaluation of scenarios for each incidents; - Development (and evaluation) of safety indicator system. The safety indicators are the most important tools used to assess the level of nuclear safety during decommissioning process, to discover the weak points and to establish safety measures. The paper contains also, a safety case evaluation (description of facility according to the decommissioning

  3. IAEA safety fundamentals: the safety of nuclear installations and the defence in depth concept

    International Nuclear Information System (INIS)

    Aro, I.

    2005-01-01

    This presentation is a replica of the similar presentation provided by the IAEA Basic Professional Training Course on Nuclear Safety. The presentation utilizes the IAEA Safety Series document No. 110, Safety Fundamentals: the Safety of Nuclear Installations. The objective of the presentation is to provide the basic rationale for actions in provision of nuclear safety. The presentation also provides basis to understand national nuclear safety requirements. There are three Safety Fundamentals documents in the IAEA Safety Series: one for nuclear safety, one for radiation safety and one for waste safety. The IAEA is currently revising its Safety Fundamentals by combining them into one general Safety Fundamentals document. The IAEA Safety Fundamentals are not binding requirements to the Member States. But, a very similar text has been provided in the Convention on Nuclear Safety which is legally binding for the Member State after ratification by the Parliament. This presentation concentrates on nuclear safety. The Safety Fundamentals documents are the 'policy documents' of the IAEA Safety Standards Series. They state the basic objectives, concepts and principles involved in ensuring protection and safety in the development and application of atomic energy for peaceful purposes. They will state - without providing technical details and without going into the application of principles - the rationale for actions necessary in meeting Safety Requirements. Chapter 7 of this presentation describes the basic features of defence in depth concept which is referred to in the Safety Fundamentals document. The defence in depth concept is a key issue in reaching high level of safety specifically at the design stage but as the reader can see the extended concept also refers to the operational stage. The appendix has been taken directly from the IAEA Basic Professional Training Course on Nuclear Safety and applied to the Finnish conditions. The text originates from the references

  4. Calculation of the BREN house shielding experiments

    International Nuclear Information System (INIS)

    Woolson, William A.; Gritzner, Michael L.

    1987-01-01

    The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these

  5. Patient safety

    African Journals Online (AJOL)

    Page 1 .... BMJ 2012;344:e832. Table 2. Unsafe medical care. Structural factors. Organisational determinants. Structural accountability (accreditation and regulation). Safety culture. Training, education and human resources. Stress and fatigue .... for routine take-off and landing, yet doctors feel that it is demeaning to do so?

  6. Sun Safety

    Science.gov (United States)

    ... Children from the Sun? Are There Benefits to Spending Time Outdoors? The Surgeon General’s Call to Action to Prevent Skin Cancer Related Resources Sun Safety Tips for Men Tips for Families Tips for Schools Tips for Employers Tips for ...

  7. Eye Involvement in TSC

    Science.gov (United States)

    ... eye involvement. Nonretinal and Retinal Eye Findings Facial angiofibromas may involve the eyelids of individuals with TSC, ... the hamartomas have many blood vessels (as are angiofibromas of the skin). Less than half of the ...

  8. A quantitative calculation for software reliability evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Jun; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    To meet these regulatory requirements, the software used in the nuclear safety field has been ensured through the development, validation, safety analysis, and quality assurance activities throughout the entire process life cycle from the planning phase to the installation phase. A variety of activities, such as the quality assurance activities are also required to improve the quality of a software. However, there are limitations to ensure that the quality is improved enough. Therefore, the effort to calculate the reliability of the software continues for a quantitative evaluation instead of a qualitative evaluation. In this paper, we propose a quantitative calculation method for the software to be used for a specific operation of the digital controller in an NPP. After injecting random faults in the internal space of a developed controller and calculating the ability to detect the injected faults using diagnostic software, we can evaluate the software reliability of a digital controller in an NPP. We tried to calculate the software reliability of the controller in an NPP using a new method that differs from a traditional method. It calculates the fault detection coverage after injecting the faults into the software memory space rather than the activity through the life cycle process. We attempt differentiation by creating a new definition of the fault, imitating the software fault using the hardware, and giving a consideration and weights for injection faults.

  9. 77 FR 8288 - Applications and Amendments to Facility Operating Licenses Involving Proposed No Significant...

    Science.gov (United States)

    2012-02-14

    ... to relief. A requestor/ petitioner who fails to satisfy these requirements with respect to at least... analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling... include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have...

  10. Criticality calculation of non-ordinary systems

    Energy Technology Data Exchange (ETDEWEB)

    Kalugin, A. V., E-mail: Kalugin-AV@nrcki.ru; Tebin, V. V. [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed.

  11. Status of the safety certification process of the TRANSRAPID system

    Energy Technology Data Exchange (ETDEWEB)

    Blomerius, J [TUEV Rheinland, Koeln (Germany). Inst. fuer Software, Elektronik, Bahntechnik

    1996-12-31

    Since 20 years TUeV Rheinland is involved in safety certification of maglev technology of the TRANSRAPID type. The process applied is called PASC (Programm Accompanying Safety Certification). The paper reports on safety assessment of relevant subsystems and components (TR07, OCS, guideway components) as well as safety certification in the final program. (HW)

  12. Issues Involving Health and Safety when Radioactive Materials are discovered

    International Nuclear Information System (INIS)

    2010-01-01

    The paper covers the contribution of man made sources of radiation to radiation dose and effects of radiation like vomiting, nausea and deaths. The late effects of radiation are cancer and leukemia with the higher the dose the higher the probability. If mutations occur in the genetic cells, effects may be inherited to the next generations.

  13. Safety Culture in activities involving ionising radiation-education project

    International Nuclear Information System (INIS)

    Sahyun, A.; Sordi, G. M.; Sanches, M. P.; Levy, P. J.; Levy, D. S.

    2004-01-01

    The Brazilian National Commission of Nuclear Energy (CNEN) requires a Radiological Protection Plan for all installations authorized to work with radioactive material and a qualified Radiological Protection Supervisor. The CNEN requires the certification of practical experience in the area plus a qualification exam, applied by them. ATOMO, has opted to develop on-line courses, using multimedia resources, available at the time this congress takes place. OMICCRON, a multimedia enterprise, is in charge of the program and design. First, the research and brochures have to be adapted for the electronic language, through links, hot words and icons, especially developed for additional information. Besides the images and graphs from the original brochures, Omiccron developed, in graphic computing, specific animation explaining the procedures in more details, illustrating and simplifying the comprehension of the more complex subjects. The CD ROM presentation was enhanced with some slide displays, automatically changing the pictures, as the explanations are given. For practical visualization of these complex and important procedures, the CD also shows some technical videos. At the end of each unit covering a specific subject, the students will be submitted to self-evaluation tests, for more profitable results. This CD is not only an electronic brochure, but mainly an on-line course with weekly Internet support via e-mail or chat site, where the learners will access the instructors and a frequent questions database, useful links and related sites, permanently upgraded. Before going to the following module, the learner has to pass a written test prepared by ATOMO, via Internet. At the end of the course, a certificate will be given. The control will be made through the use of a password, provided for the authorized company and /or users. The instructors will evaluate the learners' advancement by Internet. In the case of companies, this tool will be equally offered, by using a personal manager password. (Author) 10 refs

  14. Safety Culture in activities involving ionising radiation-education project

    Energy Technology Data Exchange (ETDEWEB)

    Sahyun, A.; Sordi, G. M.; Sanches, M. P.; Levy, P. J.; Levy, D. S.

    2004-07-01

    The Brazilian National Commission of Nuclear Energy (CNEN) requires a Radiological Protection Plan for all installations authorized to work with radioactive material and a qualified Radiological Protection Supervisor. The CNEN requires the certification of practical experience in the area plus a qualification exam, applied by them. ATOMO, has opted to develop on-line courses, using multimedia resources, available at the time this congress takes place. OMICCRON, a multimedia enterprise, is in charge of the program and design. First, the research and brochures have to be adapted for the electronic language, through links, hot words and icons, especially developed for additional information. Besides the images and graphs from the original brochures, Omiccron developed, in graphic computing, specific animation explaining the procedures in more details, illustrating and simplifying the comprehension of the more complex subjects. The CD ROM presentation was enhanced with some slide displays, automatically changing the pictures, as the explanations are given. For practical visualization of these complex and important procedures, the CD also shows some technical videos. At the end of each unit covering a specific subject, the students will be submitted to self-evaluation tests, for more profitable results. This CD is not only an electronic brochure, but mainly an on-line course with weekly Internet support via e-mail or chat site, where the learners will access the instructors and a frequent questions database, useful links and related sites, permanently upgraded. Before going to the following module, the learner has to pass a written test prepared by ATOMO, via Internet. At the end of the course, a certificate will be given. The control will be made through the use of a password, provided for the authorized company and /or users. The instructors will evaluate the learners' advancement by Internet. In the case of companies, this tool will be equally offered, by using a personal manager password. (Author) 10 refs.

  15. Effectiveness of a computer based medication calculation education and testing programme for nurses.

    Science.gov (United States)

    Sherriff, Karen; Burston, Sarah; Wallis, Marianne

    2012-01-01

    The aim of the study was to evaluate the effect of an on-line, medication calculation education and testing programme. The outcome measures were medication calculation proficiency and self efficacy. This quasi-experimental study involved the administration of questionnaires before and after nurses completed annual medication calculation testing. The study was conducted in two hospitals in south-east Queensland, Australia, which provide a variety of clinical services including obstetrics, paediatrics, ambulatory, mental health, acute and critical care and community services. Participants were registered nurses (RNs) and enrolled nurses with a medication endorsement (EN(Med)) working as clinicians (n=107). Data pertaining to success rate, number of test attempts, self-efficacy, medication calculation error rates and nurses' satisfaction with the programme were collected. Medication calculation scores at first test attempt showed improvement following one year of access to the programme. Two of the self-efficacy subscales improved over time and nurses reported satisfaction with the online programme. Results of this study may facilitate the continuation and expansion of medication calculation and administration education to improve nursing knowledge, inform practise and directly improve patient safety. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

  16. Calculations for the prediction of accident situation

    International Nuclear Information System (INIS)

    Perneczky, L.; Szabados, L.; Toth, I.

    1987-11-01

    The report deals with the analysis of the loss of feedwater transient assuming that the pressurizer safety valve remains open throughout the process. The scenario is the Paks NPP counterpart of the loss of feedwater test performed on the PMK-NVH facility. The initial and boundary conditions of the calculation agree with those of the experiment allowing direct comparison with experimental results. (author) 18 refs.; 34 figs

  17. Engineering Judgment and Natural Circulation Calculations

    OpenAIRE

    Ferreri, J. C.

    2011-01-01

    The analysis performed to establish the validity of computer code results in the particular field of natural circulation flow stability calculations is presented in the light of usual engineering practice. The effects of discretization and closure correlations are discussed and some hints to avoid undesired mistakes in the evaluations performed are given. Additionally, the results are presented for an experiment relevant to the way in which a (small) number of skilled, nuclear safety analysts...

  18. Reactor Thermal Hydraulic Numerical Calculation And Modeling

    International Nuclear Information System (INIS)

    Duong Ngoc Hai; Dang The Ba

    2008-01-01

    In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

  19. Calculational methods for lattice cells

    International Nuclear Information System (INIS)

    Askew, J.R.

    1980-01-01

    At the current stage of development, direct simulation of all the processes involved in the reactor to the degree of accuracy required is not an economic proposition, and this is achieved by progressive synthesis of models for parts of the full space/angle/energy neutron behaviour. The split between reactor and lattice calculations is one such simplification. Most reactors are constructed of repetitions of similar geometric units, the fuel elements, having broadly similar properties. Thus the provision of detailed predictions of their behaviour is an important step towards overall modelling. We shall be dealing with these lattice methods in this series of lectures, but will refer back from time to time to their relationship with overall reactor calculation The lattice cell is itself composed of somewhat similar sub-units, the fuel pins, and will itself often rely upon a further break down of modelling. Construction of a good model depends upon the identification, on physical and mathematical grounds, of the most helpful division of the calculation at this level

  20. Uncertainty calculations made easier

    International Nuclear Information System (INIS)

    Hogenbirk, A.

    1994-07-01

    The results are presented of a neutron cross section sensitivity/uncertainty analysis performed in a complicated 2D model of the NET shielding blanket design inside the ITER torus design, surrounded by the cryostat/biological shield as planned for ITER. The calculations were performed with a code system developed at ECN Petten, with which sensitivity/uncertainty calculations become relatively simple. In order to check the deterministic neutron transport calculations (performed with DORT), calculations were also performed with the Monte Carlo code MCNP. Care was taken to model the 2.0 cm wide gaps between two blanket segments, as the neutron flux behind the vacuum vessel is largely determined by neutrons streaming through these gaps. The resulting neutron flux spectra are in excellent agreement up to the end of the cryostat. It is noted, that at this position the attenuation of the neutron flux is about 1 l orders of magnitude. The uncertainty in the energy integrated flux at the beginning of the vacuum vessel and at the beginning of the cryostat was determined in the calculations. The uncertainty appears to be strongly dependent on the exact geometry: if the gaps are filled with stainless steel, the neutron spectrum changes strongly, which results in an uncertainty of 70% in the energy integrated flux at the beginning of the cryostat in the no-gap-geometry, compared to an uncertainty of only 5% in the gap-geometry. Therefore, it is essential to take into account the exact geometry in sensitivity/uncertainty calculations. Furthermore, this study shows that an improvement of the covariance data is urgently needed in order to obtain reliable estimates of the uncertainties in response parameters in neutron transport calculations. (orig./GL)

  1. Organizing Patient Involvement

    DEFF Research Database (Denmark)

    Brehm Johansen, Mette

    hospitals. During the last 25 years, patient involvement and quality improvement have become connected in Danish healthcare policy. However, the ideal of involving patients in quality improvement is described in very general terms and with only few specific expectations of how it is to be carried out...... in practice, as I show in the thesis. In the patient involvement literature, the difficulties of getting patient involvement in quality improvement to have in an impact on the planning and development of healthcare services is, for example, ascribed to conceptual vagueness of patient involvement, differences...... in perspectives, values and understandings between patients and healthcare professionals, or the lack of managerial attention and prioritization....

  2. An exercise in safety

    CERN Multimedia

    CERN Bulletin

    2014-01-01

    On 14 October, a large-scale evacuation exercise took place. Ten buildings (1-2-3-4-50-51-52-53-58-304), with a total capacity of almost 1900 people, were successfully evacuated.   The exercise, which for the first time involved all of the central buildings on the Meyrin site, was organised by the PH Department in collaboration with the HSE Unit, the GS Department and the safety officers of all the various departments involved. On the day, around 400 people were evacuated in just a few minutes.  “It took us three months to prepare for the exercise,” explains Niels Dupont, safety officer for the PH Department, who organised the exercise. “Around 100 people: safety officers, firefighters, emergency guides, observers, representatives from the control centre, etc. attended four preparatory meetings and five training sessions. We also purchased equipment such as evacuation chairs, high-visibility vests and signs to mark the evacuation route.” The dec...

  3. Plutonium safety training course

    International Nuclear Information System (INIS)

    Moe, H.J.

    1976-03-01

    This course seeks to achieve two objectives: to provide initial safety training for people just beginning work with plutonium, and to serve as a review and reference source for those already engaged in such work. Numerous references have been included to provide information sources for those wishing to pursue certain topics more fully. The first part of the course content deals with the general safety approach used in dealing with hazardous materials. Following is a discussion of the four properties of plutonium that lead to potential hazards: radioactivity, toxicity, nuclear properties, and spontaneous ignition. Next, the various hazards arising from these properties are treated. The relative hazards of both internal and external radiation sources are discussed, as well as the specific hazards when plutonium is the source. Similarly, the general hazards involved in a criticality, fire, or explosion are treated. Comments are made concerning the specific hazards when plutonium is involved. A brief summary comparison between the hazards of the transplutonium nuclides relative to 239 Pu follows. The final portion deals with control procedures with respect to contamination, internal and external exposure, nuclear safety, and fire protection. The philosophy and approach to emergency planning are also discussed

  4. Compilation of kinetic data for geochemical calculations

    International Nuclear Information System (INIS)

    Arthur, R.C.; Savage, D.; Sasamoto, Hiroshi; Shibata, Masahiro; Yui, Mikazu

    2000-01-01

    physical properties of the fracture must be homogeneous over a characteristic length that is greater than or equal to the equilibration length. If these conditions are met, calculations suggest local equilibrium would be a valid assumption in groundwater evolution models applied to the Kamaishi site if: it applies to reactions involving calcite, stilbite (assuming its dissolution / precipitation behavior is similar to that of heulandite), laumontite, albite and prehnite, but not quartz; Darcy flow velocities are relatively low (e.g., less than about 0.1 m yr-1), and it is based on the assumption that equilibrium corresponds to an uncertainty in the saturation index of 0.0±0.4. If, however, actual reaction rates in the field are lower than expected, possibly because reactive surface areas are overestimated, the modeling approach may be inappropriate because it is probably unrealistic to assume that fracture mineralogy is homogeneous over fracture lengths exceeding a few meters or tens of meters. An analytical model of redox-front migration behavior based on the stationary-state approximation, and JNC's conceptual model of a natural events scenario involving the migration of oxidizing surface waters in fractures, suggests that oxidizing solutions could travel from the surface to the depth of a repository in crystalline rock within 400 to 50,000 years. These estimates are relatively short compared with time periods considered in safety assessments of repository performance, which suggests that time-dependent variations in the redox environment of both the near field and geosphere may need to be accounted for in these assessments. The flow velocities and concentrations of reducing minerals assumed in JNC's conceptual model may be overly conservative, however. (author)

  5. Safety Assessment of Probiotics

    Science.gov (United States)

    Lahtinen, Sampo J.; Boyle, Robert J.; Margolles, Abelardo; Frias, Rafael; Gueimonde, Miguel

    Viable microbes have been a natural part of human diet throughout the history of mankind. Today, different fermented foods and other foods containing live microbes are consumed around the world, including industrialized countries, where the diet has become increasingly sterile during the last decades. By definition, probiotics are viable microbes with documented beneficial effects on host health. Probiotics have an excellent safety record, both in humans and in animals. Despite the wide and continuously increasing consumption of probiotics, adverse events related to probiotic use are extremely rare. Many popular probiotic strains such as lactobacilli and bifidobacteria can be considered as components of normal healthy intestinal microbiota, and thus are not thought to pose a risk for the host health - in contrast, beneficial effects on health are commonly reported. Nevertheless, the safety of probiotics is an important issue, in particular in the case of new potential probiotics which do not have a long history of safe use, and of probiotics belonging to species for which general assumption of safety cannot be made. Furthermore, safety of probiotics in high-risk populations such as critically ill patients and immunocompromized subjects deserves particular attention, as virtually all reported cases of bacteremia and fungemia associated with probiotic use, involve subjects with underlying diseases, compromised immune system or compromised intestinal integrity.

  6. Online plasma calculator

    Science.gov (United States)

    Wisniewski, H.; Gourdain, P.-A.

    2017-10-01

    APOLLO is an online, Linux based plasma calculator. Users can input variables that correspond to their specific plasma, such as ion and electron densities, temperatures, and external magnetic fields. The system is based on a webserver where a FastCGI protocol computes key plasma parameters including frequencies, lengths, velocities, and dimensionless numbers. FastCGI was chosen to overcome security problems caused by JAVA-based plugins. The FastCGI also speeds up calculations over PHP based systems. APOLLO is built upon the WT library, which turns any web browser into a versatile, fast graphic user interface. All values with units are expressed in SI units except temperature, which is in electron-volts. SI units were chosen over cgs units because of the gradual shift to using SI units within the plasma community. APOLLO is intended to be a fast calculator that also provides the user with the proper equations used to calculate the plasma parameters. This system is intended to be used by undergraduates taking plasma courses as well as graduate students and researchers who need a quick reference calculation.

  7. Exact and approximate multiple diffraction calculations

    International Nuclear Information System (INIS)

    Alexander, Y.; Wallace, S.J.; Sparrow, D.A.

    1976-08-01

    A three-body potential scattering problem is solved in the fixed scatterer model exactly and approximately to test the validity of commonly used assumptions of multiple scattering calculations. The model problem involves two-body amplitudes that show diffraction-like differential scattering similar to high energy hadron-nucleon amplitudes. The exact fixed scatterer calculations are compared to Glauber approximation, eikonal-expansion results and a noneikonal approximation

  8. A revised calculational model for fission

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F

    1998-09-01

    A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)

  9. Occuptional Health and Safety and Employer Motivation

    DEFF Research Database (Denmark)

    Jensen, Per Langå

    2004-01-01

    It is often argued and supported by a number of case studies that investment in human factors and occupational health and safety can pay. But any employer has a number of possible in-vestments, and many of these may have a larger marginal utility than health and safety. In addition it is often...... difficult to calculate the exact pay off for human factors and health and safety – how to calculate higher motivation for instance. The economic benefit as a possible driving force for improvement of occupational health and safety is likely to exist but it must be considered a relatively weak force. Another...... important driving force for improvements in health and safety. No employer likes to be ‘branded’ as immoral, manifested in fines by the labour inspectors or media attention to an unsafe conduct. Strategies to im-prove health and safety therefore need to focus on the legitimacy as the probably strongest...

  10. Occupational Health and Safety and Employer Motivation

    DEFF Research Database (Denmark)

    Hasle, Peter; Jensen, Per Langå

    2004-01-01

    It is often argued and supported by a number of case studies that investment in human factors and occupational health and safety can pay. But any employer has a number of possible in-vestments, and many of these may have a larger marginal utility than health and safety. In addition it is often...... difficult to calculate the exact pay off for human factors and health and safety – how to calculate higher motivation for instance. The economic benefit as a possible driving force for improvement of occupational health and safety is likely to exist but it must be considered a relatively weak force. Another...... important driving force for improvements in health and safety. No employer likes to be ‘branded’ as immoral, manifested in fines by the labour inspectors or media attention to an unsafe conduct. Strategies to im-prove health and safety therefore need to focus on the legitimacy as the probably strongest...

  11. Spent fuel storage criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E M; Elmessiry, A M [National center of nuclear safety and radiation control atomic energy authority, (Egypt)

    1995-10-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs.

  12. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

    Amin, E.M.; Elmessiry, A.M.

    1995-01-01

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs

  13. Daylight calculations in practice

    DEFF Research Database (Denmark)

    Iversen, Anne; Roy, Nicolas; Hvass, Mette

    The aim of the project was to obtain a better understanding of what daylight calculations show and also to gain knowledge of how the different daylight simulation programs perform compared with each other. Experience has shown that results for the same room, obtained from two daylight simulation...... programs can give different results. This can be due to restrictions in the program itself and/or be due to the skills of the persons setting up the models. This is crucial as daylight calculations are used to document that the demands and recommendations to daylight levels outlined by building authorities....... The aim of the project was to obtain a better understanding of what daylight calculations show and also to gain knowledge of how the different daylight simulation programs perform compared with each other. Furthermore the aim was to provide knowledge of how to build up the 3D models that were...

  14. Calculating Quenching Weights

    CERN Document Server

    Salgado, C A; Salgado, Carlos A.; Wiedemann, Urs Achim

    2003-01-01

    We calculate the probability (``quenching weight'') that a hard parton radiates an additional energy fraction due to scattering in spatially extended QCD matter. This study is based on an exact treatment of finite in-medium path length, it includes the case of a dynamically expanding medium, and it extends to the angular dependence of the medium-induced gluon radiation pattern. All calculations are done in the multiple soft scattering approximation (Baier-Dokshitzer-Mueller-Peign\\'e-Schiff--Zakharov ``BDMPS-Z''-formalism) and in the single hard scattering approximation (N=1 opacity approximation). By comparison, we establish a simple relation between transport coefficient, Debye screening mass and opacity, for which both approximations lead to comparable results. Together with this paper, a CPU-inexpensive numerical subroutine for calculating quenching weights is provided electronically. To illustrate its applications, we discuss the suppression of hadronic transverse momentum spectra in nucleus-nucleus colli...

  15. Safety Assessment in Installation of Precast Concrete

    Directory of Open Access Journals (Sweden)

    Yashrri S.N.

    2014-03-01

    Full Text Available This study was carried out to identify the safety aspects and the level of safety during the installation process in construction sites. A questionnaire survey and interviews were done to provide data on safety requirements in precast concrete construction. All of the interviews and the research questionnaire survey were conducted among contractors that are registered as class 1 to class 7 with the Construction Industry Development Board (CIDB and class A to class G with Pusat Khidmat Kontraktor (PKK in Penang. Returned questionnaires were analysed with the use of simple percentages and the Likert Scale analysis method to identify safety aspects of precast construction. The results indicate that the safety aspect implemented by companies involved in the precast construction process is at a good level in the safety aspect during bracing, propping, welding and grouting processes and at a very good level of safety in general aspects and safety aspects during lifting processes.

  16. Multimorbidity and Patient Safety Incidents in Primary Care: A Systematic Review and Meta-Analysis

    Science.gov (United States)

    Panagioti, Maria; Stokes, Jonathan; Esmail, Aneez; Coventry, Peter; Cheraghi-Sohi, Sudeh; Alam, Rahul; Bower, Peter

    2015-01-01

    Background Multimorbidity is increasingly prevalent and represents a major challenge in primary care. Patients with multimorbidity are potentially more likely to experience safety incidents due to the complexity of their needs and frequency of their interactions with health services. However, rigorous syntheses of the link between patient safety incidents and multimorbidity are not available. This review examined the relationship between multimorbidity and patient safety incidents in primary care. Methods We followed our published protocol (PROSPERO registration number: CRD42014007434). Medline, Embase and CINAHL were searched up to May 2015. Study design and quality were assessed. Odds ratios (OR) and 95% confidence intervals (95% CIs) were calculated for the associations between multimorbidity and two categories of patient safety outcomes: ‘active patient safety incidents’ (such as adverse drug events and medical complications) and ‘precursors of safety incidents’ (such as prescription errors, medication non-adherence, poor quality of care and diagnostic errors). Meta-analyses using random effects models were undertaken. Results Eighty six relevant comparisons from 75 studies were included in the analysis. Meta-analysis demonstrated that physical-mental multimorbidity was associated with an increased risk for ‘active patient safety incidents’ (OR = 2.39, 95% CI = 1.40 to 3.38) and ‘precursors of safety incidents’ (OR = 1.69, 95% CI = 1.36 to 2.03). Physical multimorbidity was associated with an increased risk for active safety incidents (OR = 1.63, 95% CI = 1.45 to 1.80) but was not associated with precursors of safety incidents (OR = 1.02, 95% CI = 0.90 to 1.13). Statistical heterogeneity was high and the methodological quality of the studies was generally low. Conclusions The association between multimorbidity and patient safety is complex, and varies by type of multimorbidity and type of safety incident. Our analyses suggest that multimorbidity

  17. Three recent TDHF calculations

    International Nuclear Information System (INIS)

    Weiss, M.S.

    1981-05-01

    Three applications of TDHF are discussed. First, vibrational spectra of a post grazing collision 40 Ca nucleus is examined and found to contain many high energy components, qualitatively consistent with recent Orsay experiments. Second, the fusion cross section in energy and angular momentum are calculated for 16 O + 24 Mg to exhibit the parameters of the low l window for this system. A sensitivity of the fusion cross section to the effective two body potential is discussed. Last, a preliminary analysis of 86 Kr + 139 La at E/sub lab/ = 505 MeV calculated in the frozen approximation is displayed, compared to experiment and discussed

  18. Fission neutron multiplicity calculations

    International Nuclear Information System (INIS)

    Maerten, H.; Ruben, A.; Seeliger, D.

    1991-01-01

    A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs

  19. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

    Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

  20. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  1. Disposal safety

    International Nuclear Information System (INIS)

    Bartlett, J.W.

    International consensus does not seem to be necessary or appropriate for many of the issues concerned with the safety of nuclear waste disposal. International interaction on the technical aspects of disposal has been extensive, and this interaction has contributed greatly to development of a consensus technical infrastructure for disposal. This infrastructure provides a common and firm base for regulatory, political, and social actions in each nation

  2. Safety aspects

    International Nuclear Information System (INIS)

    Wider, H.U.

    1997-01-01

    It is assumed that in an accelerator-driven system (ADS) the same type of accidents can be envisaged as in critical reactors. After briefly describing the basic safety features of ADS, the first investigations of the behaviour of an accelerator driven fast oxide reactor during an unprotected loss-of-flow accident and the investigation of reactivity accidents in a large sodium-cooled ADS are presented

  3. Cryogenics safety

    International Nuclear Information System (INIS)

    Reider, R.

    1977-01-01

    The safety hazards associated with handling cryogenic fluids are discussed in detail. These hazards include pressure buildup when a cryogenic fluid is heated and becomes a gas, potential damage to body tissues due to surface contact, toxic risk from breathing air altered by cryogenic fluids, dangers of air solidification, and hazards of combustible cryogens such as liquified oxygen, hydrogen, or natural gas or of combustible mixtures. Safe operating procedures and emergency planning are described

  4. Safety Checklist

    Science.gov (United States)

    1994-05-01

    given prior to issuing or renewing an OF 346? 13. Are operators’ DA Forms 348 reviewed annually for— a. Safety awards? b. Expiration of permits...place oily polishing rags or waste in covered metal cans? d. Store paint in tightly closed containers? e. Warn family members to never use gasoline...15 cream or lotion on exposed skin (face, hands, feet)? 3. Avoid extended periods of unprotected exposure to the sun? Heat cramp, heat exhaustion

  5. Nuclear safety

    International Nuclear Information System (INIS)

    2014-01-01

    The Program on Nuclear Safety comprehends Radioprotection, Radioactive Waste Management and Nuclear Material Control. These activities are developed at the Nuclear Safety Directory. The Radioactive Waste Management Department (GRR) was formally created in 1983, to promote research and development, teaching and service activities in the field of radioactive waste. Its mission is to develop and employ technologies to manage safely the radioactive wastes generated at IPEN and at its customer’s facilities all over the country, in order to protect the health and the environment of today's and future generations. The Radioprotection Service (GRP) aims primarily to establish requirements for the protection of people, as workers, contractors, students, members of the general public and the environment from harmful effects of ionizing radiation. Furthermore, it also aims to establish the primary criteria for the safety of radiation sources at IPEN and planning and preparing for response to nuclear and radiological emergencies. The procedures about the management and the control of exposures to ionizing radiation are in compliance with national standards and international recommendations. Research related to the main activities is also performed. The Nuclear Material Control has been performed by the Safeguard Service team, which manages the accountability and the control of nuclear material at IPEN facilities and provides information related to these activities to ABACC and IAEA. (author)

  6. Nuclear safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Program on Nuclear Safety comprehends Radioprotection, Radioactive Waste Management and Nuclear Material Control. These activities are developed at the Nuclear Safety Directory. The Radioactive Waste Management Department (GRR) was formally created in 1983, to promote research and development, teaching and service activities in the field of radioactive waste. Its mission is to develop and employ technologies to manage safely the radioactive wastes generated at IPEN and at its customer’s facilities all over the country, in order to protect the health and the environment of today's and future generations. The Radioprotection Service (GRP) aims primarily to establish requirements for the protection of people, as workers, contractors, students, members of the general public and the environment from harmful effects of ionizing radiation. Furthermore, it also aims to establish the primary criteria for the safety of radiation sources at IPEN and planning and preparing for response to nuclear and radiological emergencies. The procedures about the management and the control of exposures to ionizing radiation are in compliance with national standards and international recommendations. Research related to the main activities is also performed. The Nuclear Material Control has been performed by the Safeguard Service team, which manages the accountability and the control of nuclear material at IPEN facilities and provides information related to these activities to ABACC and IAEA. (author)

  7. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Shal'nov, A.V.

    1995-01-01

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  8. Evaluation of reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-04-15

    Although the operation of nuclear reactors has a remarkably good record of safety, the prevention of possible reactor accidents is one of the major factors that atomic planners have to contend with. At the same time, excessive caution may breed an attitude that hampers progress, either by resisting new development or by demanding unnecessarily elaborate and expensive precautions out of proportion to the actual hazards involved. The best course obviously is to determine the possible dangers and adopt adequate measures for their prevention, providing of course, for a reasonable margin of error in judging the hazards and the effectiveness of the measures. The greater the expert understanding and thoroughness with which this is done, the narrower need the margin be. This is the basic idea behind the evaluation of reactor safety

  9. Safety lessons from aviation.

    Science.gov (United States)

    Higton, Phil

    2005-07-01

    Thirty years ago the world of Commercial Aviation provided a challenging environment. In my early flying days, aircraft accidents were not unusual, flying was seen as a risky business and those who took part, either as a provider or passenger, appeared grudgingly willing to accept the hazards involved. A reduction in the level of risk was sought in technological advances, greater knowledge of physics and science, and access to higher levels of skill through simulation, practice and experience. While these measures did have an impact, the expected safety dividend was not realized. The most experienced, technically competent individuals with the best equipment featured far too regularly in the accident statistics. We had to look at the human element, the impact of flaws or characteristics of the human condition. We call this area Human Factors. My paper describes the concept of Human Factors, its establishment as a key safety tool in aviation and the impact of this on my working life.

  10. Explaining Ethnic Disparities in Patient Safety: A Qualitative Analysis

    NARCIS (Netherlands)

    Suurmond, Jeanine; Uiters, Ellen; de Bruijne, Martine C.; Stronks, Karien; Essink-Bot, Marie-Louise

    2010-01-01

    Objectives. We explored characteristics of in-hospital care and treatment of immigrant patients to better understand the processes underlying ethnic disparities in patient safety. Methods. We conducted semistructured interviews with care providers regarding patient safety events involving immigrant

  11. Relativistic multiple scattering X-alpha calculations

    International Nuclear Information System (INIS)

    Chermette, H.; Goursot, A.

    1986-01-01

    The necessity to include self-consistent relativistic corrections in molecular calculations has been pointed out for all compounds involving heavy atoms. Most of the changes in the electronic properties are due to the mass-velocity and the so-called Darwin terms so that the use of Wood and Boring's Hamiltonian is very convenient for this purpose as it can be easily included in MSXalpha programs. Although the spin orbit operator effects are only obtained by perturbation theory, the results compare fairly well with experiment and with other relativistic calculations, namely Hartree-Fock-Slater calculations

  12. Calculation and measurement of fog droplet size

    International Nuclear Information System (INIS)

    Laali, A.R.; Courant, J.J.; Kleitz, A.

    1991-01-01

    This paper outlines the elements involved in calculation and measurement of fog droplet size in steam turbines. The condensation calculations are performed for a 600 MW LP fossil fired, and for a 900 MW LP nuclear turbine. A simplified method based on classical condensation theory is used for these calculations. The fog droplet size measurement are carried out downstream of the last moving blades of these turbines in order to validate the program. The comparison between the results could lead to a better understanding of the condensation process in steam turbines. Some large droplet (re-entrained droplet) measurements are also taken using a microvideo probe

  13. Nuclear safety culture and nuclear safety supervision

    International Nuclear Information System (INIS)

    Chai Jianshe

    2013-01-01

    In this paper, the author reviews systematically and summarizes up the development process and stage characteristics of nuclear safety culture, analysis the connotation and characteristics of nuclear safety culture, sums up the achievements of our country's nuclear safety supervision, dissects the challenges and problems of nuclear safety supervision. This thesis focused on the relationship between nuclear safety culture and nuclear safety supervision, they are essential differences, but there is a close relationship. Nuclear safety supervision needs to introduce some concepts of nuclear safety culture, lays emphasis on humanistic care and improves its level and efficiency. Nuclear safety supervision authorities must strengthen nuclear safety culture training, conduct the development of nuclear safety culture, make sure that nuclear safety culture can play significant roles. (author)

  14. Strong-back safety latch

    International Nuclear Information System (INIS)

    DeSantis, G.N.

    1995-01-01

    The calculation decides the integrity of the safety latch that will hold the strong-back to the pump during lifting. The safety latch will be welded to the strong-back and will latch to a 1.5-in. dia cantilever rod welded to the pump baseplate. The static and dynamic analysis shows that the safety latch will hold the strong-back to the pump if the friction clamps fail and the pump become free from the strong-back. Thus, the safety latch will meet the requirements of the Lifting and Rigging Manual for under the hook lifting for static loading; it can withstand shock loads from the strong-back falling 0.25 inch

  15. Calculating Student Grades.

    Science.gov (United States)

    Allswang, John M.

    1986-01-01

    This article provides two short microcomputer gradebook programs. The programs, written in BASIC for the IBM-PC and Apple II, provide statistical information about class performance and calculate grades either on a normal distribution or based on teacher-defined break points. (JDH)

  16. Cardiovascular risk calculation

    African Journals Online (AJOL)

    James A. Ker

    2014-08-20

    Aug 20, 2014 ... smoking and elevated blood sugar levels (diabetes mellitus). These risk ... These are risk charts, e.g. FRS, a non-laboratory-based risk calculation, and ... for hard cardiovascular end-points, such as coronary death, myocardial ...

  17. Cooling tower calculations

    International Nuclear Information System (INIS)

    Simonkova, J.

    1988-01-01

    The problems are summed up of the dynamic calculation of cooling towers with forced and natural air draft. The quantities and relations are given characterizing the simultaneous exchange of momentum, heat and mass in evaporative water cooling by atmospheric air in the packings of cooling towers. The method of solution is clarified in the calculation of evaporation criteria and thermal characteristics of countercurrent and cross current cooling systems. The procedure is demonstrated of the calculation of cooling towers, and correction curves and the effect assessed of the operating mode at constant air number or constant outlet air volume flow on their course in ventilator cooling towers. In cooling towers with the natural air draft the flow unevenness is assessed of water and air relative to its effect on the resulting cooling efficiency of the towers. The calculation is demonstrated of thermal and resistance response curves and cooling curves of hydraulically unevenly loaded towers owing to the water flow rate parameter graded radially by 20% along the cross-section of the packing. Flow rate unevenness of air due to wind impact on the outlet air flow from the tower significantly affects the temperatures of cooled water in natural air draft cooling towers of a design with lower demands on aerodynamics, as early as at wind velocity of 2 m.s -1 as was demonstrated on a concrete example. (author). 11 figs., 10 refs

  18. Languages for structural calculations

    International Nuclear Information System (INIS)

    Thomas, J.B.; Chambon, M.R.

    1988-01-01

    The differences between human and computing languages are recalled. It is argued that they are to some extent structured in antagonistic ways. Languages in structural calculation, in the past, present, and future, are considered. The contribution of artificial intelligence is stressed [fr

  19. Monte Carlo alpha calculation

    Energy Technology Data Exchange (ETDEWEB)

    Brockway, D.; Soran, P.; Whalen, P.

    1985-01-01

    A Monte Carlo algorithm to efficiently calculate static alpha eigenvalues, N = ne/sup ..cap alpha..t/, for supercritical systems has been developed and tested. A direct Monte Carlo approach to calculating a static alpha is to simply follow the buildup in time of neutrons in a supercritical system and evaluate the logarithmic derivative of the neutron population with respect to time. This procedure is expensive, and the solution is very noisy and almost useless for a system near critical. The modified approach is to convert the time-dependent problem to a static ..cap alpha../sup -/eigenvalue problem and regress ..cap alpha.. on solutions of a/sup -/ k/sup -/eigenvalue problem. In practice, this procedure is much more efficient than the direct calculation, and produces much more accurate results. Because the Monte Carlo codes are intrinsically three-dimensional and use elaborate continuous-energy cross sections, this technique is now used as a standard for evaluating other calculational techniques in odd geometries or with group cross sections.

  20. Reactor dynamics calculations

    International Nuclear Information System (INIS)

    Devooght, J.; Lefvert, T.; Stankiewiez, J.

    1981-01-01

    This chapter deals with the work done in reactor dynamics within the Coordinated Research Program on Transport Theory and Advanced Reactor Calculations by three groups in Belgium, Poland, Sweden and Italy. Discretization methods in diffusion theory, collision probability methods in time-dependent neutron transport and singular perturbation method are represented in this paper

  1. Equilibrium fission model calculations

    International Nuclear Information System (INIS)

    Beckerman, M.; Blann, M.

    1976-01-01

    In order to aid in understanding the systematics of heavy ion fission and fission-like reactions in terms of the target-projectile system, bombarding energy and angular momentum, fission widths are calculated using an angular momentum dependent extension of the Bohr-Wheeler theory and particle emission widths using angular momentum coupling

  2. Evolution of calculation methods taking into account severe accidents

    International Nuclear Information System (INIS)

    L'Homme, A.; Courtaud, J.M.

    1990-12-01

    During the first decade of PWRs operation in France the calculation methods used for design and operation have improved very much. This paper gives a general analysis of the calculation methods evolution in parallel with the evolution of safety approach concerning PWRs. Then a comprehensive presentation of principal calculation tools is presented as applied during the past decade. An effort is done to predict the improvements in near future

  3. JOINT INVOLVEMENT IN SYPHILIS

    Directory of Open Access Journals (Sweden)

    T. I. Zlobina

    2016-01-01

    Full Text Available Joint involvement in syphilis has been considered as casuistry in recent years. At the same time, the high incidence of primary syphilis and the notified cases of late neurosyphilis may suggest that joint involvement in this disease is by no means always verified. Traditionally there are two forms of syphilitic arthritis: primary synovial (involving the articular membranes and sac and primary bone (involving the articular bones and cartilages ones. The paper describes the authors' clinical case of the primary bone form of articular syphilis in a 34-year-old man. 

  4. The Calculation of Flooding Level using CFX Code

    International Nuclear Information System (INIS)

    Oh, Seo Bin; Kim, Keon Yeop; Lee, Hyung Ho

    2015-01-01

    The plant design should consider internal flooding by postulated pipe ruptures, component failures, actuation of spray systems, and improper system alignment. The flooding causes failure of safety-related equipment and affects the integrity of the structure. The safety-related equipment should be installed above the flood level for protection against flooding effects. Conservative estimates of the flood level are important when a DBA occurs. The flooding level can be calculated simply applying Bernoulli's equation. However, in this study, a realistic calculation is performed with ANSYS CFX code. In calculation with CFX, air-core vortex phenomena, and turbulent flow can be simulated, which cannot be calculated analytically. The flooding level is evaluated by analytical calculation and CFX analysis for an assumed condition. The flood level is calculated as 0.71m and 1.1m analytically and with CFX simulation, respectively. Comparing the analytical calculation and simulation, they are similar, but the analytical calculation is not conservative. There are many factors reducing the drainage capacity such as air-core vortex, intake of air, and turbulent flow. Therefore, in case of flood level evaluation by analytical calculation, a sufficient safety margin should be considered

  5. Characterization strategy report for the organic safety issues

    International Nuclear Information System (INIS)

    Goheen, S.C.; Campbell, J.A.; Fryxell, G.E.

    1997-08-01

    This report describes a logical approach to resolving potential safety issues resulting from the presence of organic components in hanford tank wastes. The approach uses a structured logic diagram (SLD) to provide a pathway for quantifying organic safety issue risk. The scope of the report is limited to selected organics (i.e., solvents and complexants) that were added to the tanks and their degradation products. The greatest concern is the potential exothermic reactions that can occur between these components and oxidants, such as sodium nitrate, that are present in the waste tanks. The organic safety issue is described in a conceptual model that depicts key modes of failure-event reaction processes in tank systems and phase domains (domains are regions of the tank that have similar contents) that are depicted with the SLD. Applying this approach to quantify risk requires knowing the composition and distribution of the organic and inorganic components to determine (1) how much energy the waste would release in the various domains, (2) the toxicity of the region associated with a disruptive event, and (3) the probability of an initiating reaction. Five different characterization options are described, each providing a different level of quality in calculating the risks involved with organic safety issues. Recommendations include processing existing data through the SLD to estimate risk, developing models needed to link more complex characterization information for the purpose of estimating risk, and examining correlations between the characterization approaches for optimizing information quality while minimizing cost in estimating risk

  6. Consideration of aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Titina, B.; Cepin, M.

    2007-01-01

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

  7. Dose calculations for severe LWR accident scenarios

    International Nuclear Information System (INIS)

    Margulies, T.S.; Martin, J.A. Jr.

    1984-05-01

    This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor Accident Consequences) code. Whole body and thyroid doses are presented for a selected set of weather cases. For each weather case these calculations were performed for various times and distances including three different dose pathways - cloud (plume) shine, ground shine and inhalation. During an emergency this information can be useful since it is immediately available for projecting offsite radiological doses based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms). It can be used for emergency drill scenario development as well

  8. KENO-IV code benchmark calculation, (6)

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Naito, Yoshitaka; Yamakawa, Yasuhiro.

    1980-11-01

    A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO 3 ) 4 aqueous solution, Pu metal or PuO 2 -polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)

  9. Feasibility study on embedded transport core calculations

    International Nuclear Information System (INIS)

    Ivanov, B.; Zikatanov, L.; Ivanov, K.

    2007-01-01

    The main objective of this study is to develop an advanced core calculation methodology based on embedded diffusion and transport calculations. The scheme proposed in this work is based on embedded diffusion or SP 3 pin-by-pin local fuel assembly calculation within the framework of the Nodal Expansion Method (NEM) diffusion core calculation. The SP 3 method has gained popularity in the last 10 years as an advanced method for neutronics calculation. NEM is a multi-group nodal diffusion code developed, maintained and continuously improved at the Pennsylvania State University. The developed calculation scheme is a non-linear iteration process, which involves cross-section homogenization, on-line discontinuity factors generation, and boundary conditions evaluation by the global solution passed to the local calculation. In order to accomplish the local calculation, a new code has been developed based on the Finite Elements Method (FEM), which is capable of performing both diffusion and SP 3 calculations. The new code will be used in the framework of the NEM code in order to perform embedded pin-by-pin diffusion and SP 3 calculations on fuel assembly basis. The development of the diffusion and SP 3 FEM code is presented first following by its application to several problems. Description of the proposed embedded scheme is provided next as well as the obtained preliminary results of the C3 MOX benchmark. The results from the embedded calculations are compared with direct pin-by-pin whole core calculations in terms of accuracy and efficiency followed by conclusions made about the feasibility of the proposed embedded approach. (authors)

  10. Parent Involvement Handbook.

    Science.gov (United States)

    Caplan, Arna

    This handbook on parent involvement, designed to be used with preschool programs, was developed by the Jefferson County Public Schools in Lakewood, Colorado. Included are: (1) a general statement about parent involvement in an early childhood program, (2) a description of the Jefferson County Early Childhood Program, (3) a description of the…

  11. Trucker Safety PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2015-03-03

    This 60 second public service announcement is based on the March 2015 CDC Vital Signs report. In 2012 in the United States, about 317,000 motor vehicle crashes involved a large truck. Twenty-six thousand truck drivers and their passengers were injured in these crashes, and about 700 died. Learn what can be done to help truck drivers stay safe.  Created: 3/3/2015 by National Institute for Occupational Safety and Health (NIOSH).   Date Released: 3/3/2015.

  12. Safety training

    CERN Multimedia

    SC Unit

    2009-01-01

    Habilitation électrique A course entitled "Habilitation électrique pour personnel de laboratoire" (electrical safety qualification for laboratory personnel) will be held on 22 and 23 June. Registration by e-mail to isabelle.cusato@cern.ch. Explosion Hazards in the handling of flammable solvents and gases A course entitled "Explosion Hazards in the handling of flammable solvents and gases" given in French will be held on 18-19 June 2009. This course is obligatory for all FGSOs at CERN, and it is recommended for anyone handling flammable gas or solvents. To sign up please visit this page. For more information please contact Isabelle Cusato, tel. 73811.

  13. SAFETY NOTES

    CERN Document Server

    TIS Secretariat

    2001-01-01

    Please note that the revisions of safety notes no 3 (NS 3 Rev. 2) and no 24 (NS 24 REV.) entitled respectively 'FIRE PREVENTION FOR ENCLOSED SPACES IN LARGE HALLS' and 'REMOVING UNBURIED ELV AND LVA ELECTRIC CONDUITS' are available on the web at the following urls: http://edmsoraweb.cern.ch:8001/cedar/doc.download?document_id=322811&version=1&filename=version_francaise.pdf http://edmsoraweb.cern.ch:8001/cedar/doc.download?document_id=322861&version=2&filename=version_francaise.pdf Paper copies can also be obtained from the TIS Divisional Secretariat, email tis.secretariat@cern.ch

  14. Course on hybrid calculation

    International Nuclear Information System (INIS)

    Weill, J.; Tellier; Bonnemay; Craigne; Chareton; Di Falco

    1969-02-01

    After a definition of hybrid calculation (combination of analogue and digital calculation) with a distinction between series and parallel hybrid computing, and a description of a hybrid computer structure and of task sharing between computers, this course proposes a description of hybrid hardware used in Saclay and Cadarache computing centres, and of operations performed by these systems. The next part addresses issues related to programming languages and software. The fourth part describes how a problem is organised for its processing on these computers. Methods of hybrid analysis are then addressed: resolution of optimisation problems, of partial differential equations, and of integral equations by means of different methods (gradient, maximum principle, characteristics, functional approximation, time slicing, Monte Carlo, Neumann iteration, Fischer iteration)

  15. Calculation of projected ranges

    International Nuclear Information System (INIS)

    Biersack, J.P.

    1980-09-01

    The concept of multiple scattering is reconsidered for obtaining the directional spreading of ion motion as a function of energy loss. From this the mean projection of each pathlength element of the ion trajectory is derived which - upon summation or integration - leads to the desired mean projected range. In special cases, the calculation can be carried out analytically, otherwise a simple general algorithm is derived which is suitable even for the smallest programmable calculators. Necessary input for the present treatment consists only of generally accessable stopping power and straggling formulas. The procedure does not rely on scattering cross sections, e.g. power potential or f(t 1 sup(/) 2 ) approximations. The present approach lends itself easily to include electronic straggling or to treat composed target materials, or even to account for the so-called time integral. (orig.)

  16. Calculation of polarization effects

    International Nuclear Information System (INIS)

    Chao, A.W.

    1983-09-01

    Basically there are two areas of accelerator applications that involve beam polarization. One is the acceleration of a polarized beam (most likely a proton beam) in a synchrotron. Another concerns polarized beams in an electron storage ring. In both areas, numerical techniques have been very useful

  17. Spallation reactions: calculations

    International Nuclear Information System (INIS)

    Bertini, H.W.

    1975-01-01

    Current methods for calculating spallation reactions over various energy ranges are described and evaluated. Recent semiempirical fits to existing data will probably yield the most accurate predictions for these reactions in general. However, if the products in question have binding energies appreciably different from their isotropic neighbors and if the cross section is approximately 30 mb or larger, then the intranuclear-cascade-evaporation approach is probably better suited. (6 tables, 12 figures, 34 references) (U.S.)

  18. Mobile Phone Network Operators' Actions on RF Safety (invited paper)

    International Nuclear Information System (INIS)

    Causebrook, J.H.

    1999-01-01

    The current and possible future global penetration of mobile phone usage is given. Health and safety aspects relate to both the statutory requirements for the operation of their networks and the public perception of risks in using services provided by the operators. The coordination of this work nationally through trade associations is mentioned. GSM is the predominant standard used for the provision of global mobile phone services. The GSM MoU Association is introduced as the operators' coordination body worldwide for dealing with radio frequency (RF) health and safety issues through its sub-group, EBRC. The scope of the EBRC group is presented with the considerations used to determine if external research should be supported by the GSM MoU Association. A personal view is provided on the present quality of worldwide research on RF health and safety and some consideration is given as to what constitutes 'good' research. The mobile phone network operators' involvement in the science and application of epidemiological research is considered. Consideration is given to introducing risk/benefit analysis into the debate on the health and safety of mobile phone usage. The media presentation of the results of scientific work on this topic often leads to a falsely negative public perception of the perceived risks. This is made worse when such perceptions are used for the purposes of objecting to the deployment of network infrastructure. The operators' approach to RF health and safety procedures is outlined, with a clarification of the distinctions between near-field and far-field methodologies for the calculation of physical exclusion zones. It is concluded that the mobile phone operators are part of an industry which is safe and who work to ensure that their operations are seen to be safe in the context of the best available worldwide scientific knowledge and safety guidelines. (author)

  19. Performance assessment calculational exercises

    International Nuclear Information System (INIS)

    Barnard, R.W.; Dockery, H.A.

    1990-01-01

    The Performance Assessment Calculational Exercises (PACE) are an ongoing effort coordinated by Yucca Mountain Project Office. The objectives of fiscal year 1990 work, termed PACE-90, as outlined in the Department of Energy Performance Assessment (PA) Implementation Plan were to develop PA capabilities among Yucca Mountain Project (YMP) participants by calculating performance of a Yucca Mountain (YM) repository under ''expected'' and also ''disturbed'' conditions, to identify critical elements and processes necessary to assess the performance of YM, and to perform sensitivity studies on key parameters. It was expected that the PACE problems would aid in development of conceptual models and eventual evaluation of site data. The PACE-90 participants calculated transport of a selected set of radionuclides through a portion of Yucca Mountain for a period of 100,000 years. Results include analyses of fluid-flow profiles, development of a source term for radionuclide release, and simulations of contaminant transport in the fluid-flow field. Later work included development of a problem definition for perturbations to the originally modeled conditions and for some parametric sensitivity studies. 3 refs

  20. A Tabular Approach to Titration Calculations

    Science.gov (United States)

    Lim, Kieran F.

    2012-01-01

    Titrations are common laboratory exercises in high school and university chemistry courses, because they are easy, relatively inexpensive, and they illustrate a number of fundamental chemical principles. While students have little difficulty with calculations involving a single titration step, there is a significant leap in conceptual difficulty…