WorldWideScience

Sample records for safety calculations involving

  1. 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...

  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. 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

  4. 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.

  5. 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.

  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. 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

  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. 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.

  10. 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.

  11. 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)

  12. 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

  13. 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

  14. 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.)

  15. 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)

  16. 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

  17. 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...

  18. 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...

  19. 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...

  20. 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...

  1. 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.

  2. 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.

  3. 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

  4. 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)

  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. 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.

  7. 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...

  8. 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

  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. 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

  11. 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

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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)

  5. 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

  6. 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...

  7. 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

  8. 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.

  9. 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.).

  10. 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

  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. 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.

  13. : 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...

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. 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.)

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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...

  8. 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.)

  9. 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.)

  10. 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...

  11. 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)

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.)

  18. 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)

  19. 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)

  20. 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

  1. 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

  2. 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

  3. [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.

  4. 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.)

  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. 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

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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

  13. 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

  14. 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.

  15. 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)

  16. 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.

  17. 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

  18. 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)

  19. 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)

  20. 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)

  1. 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%.

  2. 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)

  3. 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)

  4. 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...

  5. 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

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  12. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

    2008-12-01

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

  13. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  14. 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)

  15. 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.

  16. Exploring relationships between hospital patient safety culture and Consumer Reports safety scores.

    Science.gov (United States)

    Smith, Scott Alan; Yount, Naomi; Sorra, Joann

    2017-02-16

    A number of private and public companies calculate and publish proprietary hospital patient safety scores based on publicly available quality measures initially reported by the U.S. federal government. This study examines whether patient safety culture perceptions of U.S. hospital staff in a large national survey are related to publicly reported patient safety ratings of hospitals. The Agency for Healthcare Research and Quality Hospital Survey on Patient Safety Culture (Hospital SOPS) assesses provider and staff perceptions of hospital patient safety culture. Consumer Reports (CR), a U.S. based non-profit organization, calculates and shares with its subscribers a Hospital Safety Score calculated annually from patient experience survey data and outcomes data gathered from federal databases. Linking data collected during similar time periods, we analyzed relationships between staff perceptions of patient safety culture composites and the CR Hospital Safety Score and its five components using multiple multivariate linear regressions. We analyzed data from 164 hospitals, with patient safety culture survey responses from 140,316 providers and staff, with an average of 856 completed surveys per hospital and an average response rate per hospital of 56%. Higher overall Hospital SOPS composite average scores were significantly associated with higher overall CR Hospital Safety Scores (β = 0.24, p Consumer Reports Hospital Safety Score, which is a composite of patient experience and outcomes data from federal databases. As hospital managers allocate resources to improve patient safety culture within their organizations, their efforts may also indirectly improve consumer-focused, publicly reported hospital rating scores like the Consumer Reports Hospital Safety Score.

  17. 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.

  18. 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...

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. Project Guarantee 1985. Final repository for high-level radioactive wastes: Safety report

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Disposal of radioactive was involves preventing releases to the biosphere for a long period of time and subsequently limiting the magnitude of releases by means of a series of safety barriers: the waste solidification matrix (borosilicate glass), massive steel canisters in highly compacted bentonite, sealing of void spacer and access routes on repository closure. The geological barriers are formed by the crystalline bed-rock and the overlying sedimentary layers. In order to perform a safety assessment the behaviour of these technical barriers and of the host rock must be understood and this understanding must be translated into quantitative models which allow calculation of repository performance. For the particular case of a Swiss repository, the main criterion is the individual dose limit of 10 mrem/year, which is given in the safety guidelines of the Swiss authorities. The procedure for the safety analysis involves examination of all scenarios which could give rise to radionuclide release from the repository. Qualitative considerations of both the magnitude of their consequences and their likelihood are used in order to identify a restricted number of scenarios for quantitative analysis

  8. 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

  9. International Handbook of Evaluated Criticality Safety Benchmark Experiments - ICSBEP (DVD), Version 2013

    International Nuclear Information System (INIS)

    2013-01-01

    The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 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) became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical experiment 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 and to determine criticality alarm requirement 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 nearly 66,000 pages and contain 558 evaluations with benchmark specifications for 4,798 critical, near critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the Handbook are benchmark specifications for Critical, Bare, HEU(93.2)- Metal Sphere experiments referred to as ORSphere that were performed by a team of experimenters at Oak Ridge National Laboratory in the early 1970's. A photograph of this assembly is shown on the front cover

  10. 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)

  11. 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.

  12. 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

  13. 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%

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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...

  2. 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…

  3. Safety behavior: Job demands, job resources, and perceived management commitment to safety.

    Science.gov (United States)

    Hansez, Isabelle; Chmiel, Nik

    2010-07-01

    The job demands-resources model posits that job demands and resources influence outcomes through job strain and work engagement processes. We test whether the model can be extended to effort-related "routine" safety violations and "situational" safety violations provoked by the organization. In addition we test more directly the involvement of job strain than previous studies which have used burnout measures. Structural equation modeling provided, for the first time, evidence of predicted relationships between job strain and "routine" violations and work engagement with "routine" and "situational" violations, thereby supporting the extension of the job demands-resources model to safety behaviors. In addition our results showed that a key safety-specific construct 'perceived management commitment to safety' added to the explanatory power of the job demands-resources model. A predicted path from job resources to perceived management commitment to safety was highly significant, supporting the view that job resources can influence safety behavior through both general motivational involvement in work (work engagement) and through safety-specific processes.

  4. 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.

  5. 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

  6. 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.)

  7. Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

    International Nuclear Information System (INIS)

    Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

    1993-01-01

    Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking

  8. 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.

  9. 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

  10. 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

  11. 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

  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. 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)

  14. Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F.; Santos, Thiago Augusto dos

    2011-01-01

    This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)

  15. 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 ...

  16. 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),...

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. DOE spent nuclear fuel -- Nuclear criticality safety challenges and safeguards initiatives

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1994-01-01

    The field of nuclear criticality safety is confronted with growing technical challenges and the need for forward-thinking initiatives to address and resolve issues surrounding economic, safe and secure packaging, transport, interim storage, and long-term disposal of spent nuclear fuel. These challenges are reflected in multiparameter problems involving optimization of packaging designs for maximizing the density of material per package while ensuring subcriticality and safety under variable normal and hypothetical transport and storage conditions and for minimizing costs. Historic and recently revealed uncertainties in basic data used for performing nuclear subcriticality evaluations and safety analyses highlight the need to be vigilant in assessing the validity and range of applicability of calculational evaluations that represent extrapolations from ''benchmark'' data. Examples of these uncertainties are provided. Additionally, uncertainties resulting from the safeguarding of various forms of fissionable materials in transit and storage are discussed

  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. 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

  6. 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.

  7. 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

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  14. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  15. 2005 dossier: granite. Tome: safety analysis of the geologic disposal

    International Nuclear Information System (INIS)

    2005-01-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

  16. 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.

  17. Statistical considerations on safety analysis

    International Nuclear Information System (INIS)

    Pal, L.; Makai, M.

    2004-01-01

    The authors have investigated the statistical methods applied to safety analysis of nuclear reactors and arrived at alarming conclusions: a series of calculations with the generally appreciated safety code ATHLET were carried out to ascertain the stability of the results against input uncertainties in a simple experimental situation. Scrutinizing those calculations, we came to the conclusion that the ATHLET results may exhibit chaotic behavior. A further conclusion is that the technological limits are incorrectly set when the output variables are correlated. Another formerly unnoticed conclusion of the previous ATHLET calculations that certain innocent looking parameters (like wall roughness factor, the number of bubbles per unit volume, the number of droplets per unit volume) can influence considerably such output parameters as water levels. The authors are concerned with the statistical foundation of present day safety analysis practices and can only hope that their own misjudgment will be dispelled. Until then, the authors suggest applying correct statistical methods in safety analysis even if it makes the analysis more expensive. It would be desirable to continue exploring the role of internal parameters (wall roughness factor, steam-water surface in thermal hydraulics codes, homogenization methods in neutronics codes) in system safety codes and to study their effects on the analysis. In the validation and verification process of a code one carries out a series of computations. The input data are not precisely determined because measured data have an error, calculated data are often obtained from a more or less accurate model. Some users of large codes are content with comparing the nominal output obtained from the nominal input, whereas all the possible inputs should be taken into account when judging safety. At the same time, any statement concerning safety must be aleatory, and its merit can be judged only when the probability is known with which the

  18. 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.

  19. 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.

  20. 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.

  1. Commissioning of research reactors. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

  2. 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

  3. 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

  4. International handbook of evaluated criticality safety benchmark experiments

    International Nuclear Information System (INIS)

    2010-01-01

    The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 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) became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments 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 and to determine criticality alarm requirement 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 over 55,000 pages and contain 516 evaluations with benchmark specifications for 4,405 critical, near critical, or subcritical configurations, 24 criticality alarm placement / shielding configurations with multiple dose points for each, and 200 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these evaluations; however, benchmark specifications are not derived for such experiments (in some cases models are provided in an appendix). Approximately 770 experimental configurations are categorized as unacceptable for use as criticality safety benchmark experiments. Additional evaluations are in progress and will be

  5. Stakeholder involvement in international conventions governing civil nuclear activities

    International Nuclear Information System (INIS)

    Emmerechts, Sam

    2017-01-01

    Mr Emmerechts explained that international conventions have varying positions on stakeholders and their involvement depending upon the intent of the legislator and the field they cover, ranging from a narrow to a broad interpretation. He addressed stakeholder involvement in two other international conventions governing civil nuclear activities, namely the Convention on Nuclear Safety, and the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (the Joint Convention), both concluded under the auspices of the International Atomic Energy Agency (IAEA). He noted that the Convention on Nuclear Safety remains a 'traditional' international legal instrument, focusing on governments and governmental bodies as the main stakeholders and limiting obligations regarding the involvement of the public and intergovernmental organisations to their receiving information and observing. Likewise, the Joint Convention limits obligations regarding public involvement to access to information, notably as to the siting of proposed facilities. However, he noted that in the European Union, the Directive on Nuclear Safety (2014/87/Euratom) and the Directive for the Safe Management of Spent Fuel and Radioactive Waste (2011/70/Euratom) have more advanced public participation requirements in nuclear decision making. Mr Emmerechts explained that the substantial differences between nuclear legislation and the Aarhus and Espoo Conventions with regards to public involvement requirements could partly be explained by the technicality of nuclear information and by issues related to nuclear security

  6. 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).

  7. 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)

  8. Regulatory considerations for computational requirements for nuclear criticality safety

    International Nuclear Information System (INIS)

    Bidinger, G.H.

    1995-01-01

    As part of its safety mission, the U.S. Nuclear Regulatory Commission (NRC) approves the use of computational methods as part of the demonstration of nuclear criticality safety. While each NRC office has different criteria for accepting computational methods for nuclear criticality safety results, the Office of Nuclear Materials Safety and Safeguards (NMSS) approves the use of specific computational methods and methodologies for nuclear criticality safety analyses by specific companies (licensees or consultants). By contrast, the Office of Nuclear Reactor Regulation approves codes for general use. Historically, computational methods progressed from empirical methods to one-dimensional diffusion and discrete ordinates transport calculations and then to three-dimensional Monte Carlo transport calculations. With the advent of faster computational ability, three-dimensional diffusion and discrete ordinates transport calculations are gaining favor. With the proper user controls, NMSS has accepted any and all of these methods for demonstrations of nuclear criticality safety

  9. 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

  10. 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.

  11. Modular reliability modeling of the TJNAF personnel safety system

    International Nuclear Information System (INIS)

    Cinnamon, J.; Mahoney, K.

    1997-01-01

    A reliability model for the Thomas Jefferson National Accelerator Facility (formerly CEBAF) personnel safety system has been developed. The model, which was implemented using an Excel spreadsheet, allows simulation of all or parts of the system. Modularity os the model's implementation allows rapid open-quotes what if open-quotes case studies to simulate change in safety system parameters such as redundancy, diversity, and failure rates. Particular emphasis is given to the prediction of failure modes which would result in the failure of both of the redundant safety interlock systems. In addition to the calculation of the predicted reliability of the safety system, the model also calculates availability of the same system. Such calculations allow the user to make tradeoff studies between reliability and availability, and to target resources to improving those parts of the system which would most benefit from redesign or upgrade. The model includes calculated, manufacturer's data, and Jefferson Lab field data. This paper describes the model, methods used, and comparison of calculated to actual data for the Jefferson Lab personnel safety system. Examples are given to illustrate the model's utility and ease of use

  12. 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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. Safety prediction for basic components of safety-critical software based on static testing

    International Nuclear Information System (INIS)

    Son, H.S.; Seong, P.H.

    2000-01-01

    The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

  15. 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....

  16. A consistent approach to assess safety criteria for reactivity initiated accidents

    International Nuclear Information System (INIS)

    Sartoris, C.; Taisne, A.; Petit, M.; Barre, F.; Marchand, O.

    2010-01-01

    In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on: -A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena; -The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results; -The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up. This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO 2 rod at Hot Zero Power.

  17. 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

  18. 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.

  19. Thermophysical aspects of WWER safety

    International Nuclear Information System (INIS)

    Kolochko, Vladimir N.

    1999-01-01

    The paper presents a review of the main thermophysical aspects of NPPs safety and efficiency increase applied to WWERs. Improvement of operating WWER units is the main short-term and medium-term tasks of the utilities in Ukraine. The new generation of reactors for increasing the reactor facilities efficiency should utilize achievements of thermo physics research results. The thermophysical aspects of NPPs safety and efficiency are envisaged in the context of the atomic energy development strategy. The analysis of thermophysical processes occurring in the core shows that a number of problems concerning boiling crisis and heat exchange intensification during transient and accidents are not solved in spite of numerous calculations and experimental research. The up-to-date safety conception includes severe accidents consideration for safety assessment. Review of severe accidents management is presented. Additional validation of the West advanced thermal hydraulic codes which are applied for WWERs reassessment calculations is required. Contribution of the processes which might occur in WWER containment to safety problems solving is considered as well. (author)

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

    International Nuclear Information System (INIS)

    Choi, Y. S.; Choi, K. S.; Choi, K. W.; Song, I. J.; Park, D. K.

    2001-01-01

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

  1. Safety prediction for basic components of safety critical software based on static testing

    International Nuclear Information System (INIS)

    Son, H.S.; Seong, P.H.

    2001-01-01

    The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, both of which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

  2. 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

  3. Implications of Monte Carlo Statistical Errors in Criticality Safety Assessments

    International Nuclear Information System (INIS)

    Pevey, Ronald E.

    2005-01-01

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

  4. 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

  5. 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.

  6. 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...

  7. 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...

  8. Need for an "integrated safety assessment" of GMOs, linking food safety and environmental considerations.

    Science.gov (United States)

    Haslberger, Alexander G

    2006-05-03

    Evidence for substantial environmental influences on health and food safety comes from work with environmental health indicators which show that agroenvironmental practices have direct and indirect effects on human health, concluding that "the quality of the environment influences the quality and safety of foods" [Fennema, O. Environ. Health Perspect. 1990, 86, 229-232). In the field of genetically modified organisms (GMOs), Codex principles have been established for the assessment of GM food safety and the Cartagena Protocol on Biosafety outlines international principles for an environmental assessment of living modified organisms. Both concepts also contain starting points for an assessment of health/food safety effects of GMOs in cases when the environment is involved in the chain of events that could lead to hazards. The environment can act as a route of unintentional entry of GMOs into the food supply, such as in the case of gene flow via pollen or seeds from GM crops, but the environment can also be involved in changes of GMO-induced agricultural practices with relevance for health/food safety. Examples for this include potential regional changes of pesticide uses and reduction in pesticide poisonings resulting from the use of Bt crops or influences on immune responses via cross-reactivity. Clearly, modern methods of biotechnology in breeding are involved in the reasons behind the rapid reduction of local varieties in agrodiversity, which constitute an identified hazard for food safety and food security. The health/food safety assessment of GM foods in cases when the environment is involved needs to be informed by data from environmental assessment. Such data might be especially important for hazard identification and exposure assessment. International organizations working in these areas will very likely be needed to initiate and enable cooperation between those institutions responsible for the different assessments, as well as for exchange and analysis of

  9. State of the art of CATHARE model for transient safety analysis of ASTRID SFR

    International Nuclear Information System (INIS)

    Lavastre, R.; Conti, A.; Marsault, Ph.; Chenaud, M.S.; Tosello, A.

    2014-01-01

    Within the framework of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), the conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves enhancing the general design in order to : - increase the safety margins for all unprotected-loss-of-flow (ULOF) and unprotected-loss-of-heat-sink (ULOHS) transients, - identify the need for additional safety devices that would complement core natural behavior so that temperature criteria on coolant, core and primary circuit structures can remain under the safety criteria. For this purpose, the use of CATHARE system code has been very important from the early stage of design in order to ensure a feedback for design teams to improve behavior during unprotected transients. Until 2012, CATHARE ULOxx transient calculations have been used mainly to compare different core designs. They contributed to lead to the choice of CFV core (axially heterogeneous core with an upper sodium plenum employed to achieve a negative sodium void reactivity worth). Meanwhile, models for an accurate core description and transients have been developed in CATHARE to improve the calculations towards best estimate calculations for safety analysis. This paper therefore presents these main developments in core modeling achieved for the 2 past years. For instance, we will focus on the way of dealing with fuel assemblies that have to be grouped together in the CATHARE code to form a channel with similar neutronic physics and thermal-hydraulics characteristics. We will also explain the way we deal with heterogeneity of fuel pin to obtain the accurate fuel temperature along the axis and to take into account pellet-cladding gap state. These two points have a great importance on feedback effects linked to the fuel, mainly the Doppler effect. The paper will finally introduce the upcoming improvements that are under development nowadays

  10. University of New Mexico short course in nuclear criticality safety: Training for new NCS [nuclear criticality safety] specialists

    International Nuclear Information System (INIS)

    Busch, R.D.

    1990-01-01

    Since 1973, the University of New Mexico (UNM) has given ten short courses in nuclear criticality safety (NCS). Generally, thee have been given every other year, although in 1989 it was decided to offer the course on an annual basis. This decision was primarily based on the large demand for NCS specialists and a large turnover rate in the industry. The purpose of the course is to provide a 1-week overview of NCS. The typical student has been involved in NCS for <1 yr, although it many cases they have been associated with the nuclear industry in other capacities for many years. The short course is conducted at several levels. Carefully prepared lectures provide the information framework for selected topics. The following topics are covered in the course: basic reactor theory, criticality accidents and consequences, hand calculations, administration of a criticality safety program, regulators and their processes, computer methods and applications, experimental methods and correlations, overview of some process operations, and transportation and storage issues in NCS

  11. 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

  12. 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...

  13. Safety guide data on radiation shielding in a reprocessing facility

    International Nuclear Information System (INIS)

    Sekiguchi, Noboru; Naito, Yoshitaka

    1986-04-01

    In a reprocessing facility, various radiation sources are handled and have many geometrical conditions. To aim drawing up a safety guidebook on radiation shielding in order to evaluate shielding safety in a reprocessing facility with high reliability and reasonableness, JAERI trusted investigation on safety evaluation techniques of radiation shielding in a reprocessing facility to Nuclear Safety Research Association. This report is the collection of investigation results, and describes concept of shielding safety design principle, radiation sources in reprocessing facility and estimation of its strength, techniques of shielding calculations, and definite examples of shielding calculation in reprocessing facility. (author)

  14. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de.

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  15. 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

  16. 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

  17. 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...

  18. 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

  19. Thermal Margin Calculation of the CAREM-25 Core

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

    During the operation in steady state and anticipated operational transient of a nuclear reactor it is necessary to avoid the damage in the fuel elements induced by thermal or hydraulic effects.To satisfy that design bases safety limits are established and calculation methodologies are defined to verify them.In the particular case of the reactor CAREM-25 reactor where the core is cooled by natural circulation, it is not adequate to use directly the same calculation methodologies from typical PWR and BWR.The low cooling flow rate and not having channels in the fuel elements (open-channel fuels) produce that most of the models and computer programs typically used must be carefully validated.As result of that process, an adequate calculation procedure for this reactor type was developed.In the present work, the thermal-hydraulic design criteria of the core and the design bases, the uncertainties factors, and the thermal margin results of the core are described.Despite that the methodology of DNBR calculation is under a validation process and considering the available calculation tools, it is possible to assure that the core fulfills the safety regulations in steady state conditions

  20. A management system integrating radiation protection and safety supporting safety culture in the hospital

    International Nuclear Information System (INIS)

    Almen, A.; Lundh, C.

    2015-01-01

    Quality assurance has been identified as an important part of radiation protection and safety for a considerable time period. A rational expansion and improvement of quality assurance is to integrate radiation protection and safety in a management system. The aim of this study was to explore factors influencing the implementing strategy when introducing a management system including radiation protection and safety in hospitals and to outline benefits of such a system. The main experience from developing a management system is that it is possible to create a vast number of common policies and routines for the whole hospital, resulting in a cost-efficient system. One of the key benefits is the involvement of management at all levels, including the hospital director. Furthermore, a transparent system will involve staff throughout the organisation as well. A management system supports a common view on what should be done, who should do it and how the activities are reviewed. An integrated management system for radiation protection and safety includes key elements supporting a safety culture. (authors)

  1. 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)

  2. SKI's and SSI's review of SKB's safety report SR-Can

    International Nuclear Information System (INIS)

    Dverstorp, Bjoern; Stroemberg, Bo

    2008-03-01

    This report summarises SKI's and SSI's joint review of the Swedish Nuclear Fuel and Waste Management Co's (SKB) safety report SR-Can (SKB TR-06-09). SR-Can is the first assessment of post-closure safety for a KBS-3 spent nuclear fuel repository at the candidate sites Forsmark and Laxemar, respectively. The analysis builds on data from the initial stage of SKB's surface-based site investigations and on data from full-scale manufacturing and testing of buffer and copper canisters. SR-Can can be regarded as a preliminary version of the safety report that will be required in connection with SKB's planned licence application for a final repository in late 2009. The main purpose of the authorities' review is to provide feedback to SKB on their safety reporting as part of the pre-licensing consultation process. However, SR-Can is not part of the formal licensing process. In support of the authorities' review three international peer review teams were set up to make independent reviews of SR-Can from three perspectives, namely integration of site data, representation of the engineered barriers and safety assessment methodology, respectively. Further, several external experts and consultants have been engaged to review detailed technical and scientific issues in SR-Can. The municipalities of Oesthammar and Oskarshamn where SKB is conducting site investigations, as well NGOs involved in SKB's programme, have been invited to provide their views on SR-Can as input to the authorities' review. Finally, the authorities themselves, and with the help of consultants, have used independent models to reproduce part of SKB's calculations and to make complementary calculations. All supporting review documents are published in SKI's and SSI's report series. The main findings of the review are: -SKB's safety assessment methodology is overall in accordance with applicable regulations, but part of the methodology needs to be further developed for the licence application. -SKB's quality

  3. ATHLET calculations of the pressurizer surge line break (PH-SLB test) at the PMK-2 test facility

    International Nuclear Information System (INIS)

    Krepper, E.; Schaefer, F.

    2000-01-01

    At the Hungarian integral test facility PMK-2 a pressurizer surge line break experiment (PH-SLB test) was carried out with the PHARE 4.2.6b project. The primary objective of the test was to provide experimental data for a surge line break transient at VVER-440 reactors with reduced injection from the emergency core cooling systems (ECC). At the Institute of Safety Research calculations of the experiment were performed with the thermohydraulic computer code ATHLET, which was developed by GRS (Gesellschaft fuer Anlagen- und Reaktorsicherheit) mbH. In the context of the PHARE 4.2.6b project the Institute of Safety Research has also supplied the void fraction measurement system for the PMK-2 test facility and was involved in the evaluation of the experimental results. (orig.)

  4. 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.

  5. 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.)

  6. Safety functions and safety function indicators - key elements in SKB'S methodology for assessing long-term safety of a KBS-3 repository

    International Nuclear Information System (INIS)

    Hedin, A.

    2008-01-01

    The application of so called safety function indicators in SKB safety assessment of a KBS-3 repository for spent nuclear fuel is presented. Isolation and retardation are the two main safety functions of the KBS-3 concept. In order to quantitatively evaluate safety on a sub-system level, these functions need to be differentiated, associated with quantitative measures and, where possible, with quantitative criteria relating to the fulfillment of the safety functions. A safety function is defined as a role through which a repository component contributes to safety. A safety function indicator is a measurable or calculable property of a repository component that allows quantitative evaluation of a safety function. A safety function indicator criterion is a quantitative limit such that if the criterion is fulfilled, the corresponding safety function is upheld. The safety functions and their associated indicators and criteria developed for the KBS-3 repository are primarily related to the isolating potential and to physical states of the canister and the clay buffer surrounding the canister. They are thus not directly related to release rates of radionuclides. The paper also describes how the concepts introduced i) aid in focussing the assessment on critical, safety related issues, ii) provide a framework for the accounting of safety throughout the different time frames of the assessment and iii) provide key information in the selection of scenarios for the safety assessment. (author)

  7. 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)

  8. 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

  9. Calculations in support of the MNR core conversion

    International Nuclear Information System (INIS)

    Day, S.E.; Butler, M.P.; Garland, Wm. J.

    2002-01-01

    Calculations and results in support of the HEU to LEU fuel conversion for the McMaster Nuclear Reactor are described. Static reactor physics studies were used to determine local and global power distributions; facilitating the definition of a Reference Core configuration for mixed HEU-LEU and complete LEU loadings. Fission product inventory calculations were used to compare the two fuel enrichments from a radiological hazard point of view. Thermalhydraulic models were created and analyzed to determine steady-state temperature distributions and safety margins, and used as a scoping tool the in development of a full core thermalhydraulic model. The behaviour of the two enrichment fuels was investigated in the context of a protected startup transient. The simulation results support the conclusion that the LEU fuel behaves in much the same way as the HEU fuel, which it is replacing. The conversion results in no new safety issues or significant changes in safety parameters. (author)

  10. Researches on nuclear criticality safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-10-01

    For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

  11. Researches on nuclear criticality safety evaluation

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi

    2003-01-01

    For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

  12. 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

  13. 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

  14. Safety Training Parks – Cooperative Contribution to Safety and Health Trainings

    DEFF Research Database (Denmark)

    Reiman, Arto; Pedersen, Louise Møller; Väyrynen, Seppo

    2017-01-01

    . The concept of Safety Training Park (STP) has been developed to meet these challenges. Eighty stakeholders from the Finnish construction industry have been involved in the construction and financing of the STP in northern Finland (STPNF). This unique cooperation has contributed to the immediate success......, and evidence from the literature are presented with a focus on the pros and cons of the STPNF. The STP is a new and innovative method for safety training that stimulates different learning styles and inspires changes in individuals’ behavior and in the organizations’ safety climate. The stakeholders’ high...... commitment, a long-term perspective, and a strong safety climate are identified as preconditions for the STP concept to work....

  15. Nuclear critical safety analysis for UX-30 transport of freight package

    International Nuclear Information System (INIS)

    Quan Yanhui; Zhou Qi; Yin Shenggui

    2014-01-01

    The nuclear critical safety analysis and evaluation for UX-30 transport freight package in the natural condition and accident condition were carried out with MONK-9A code and MCNP code. Firstly, the critical benchmark experiment data of public in international were selected, and the deflection and subcritical limiting value with MONK-9A code and MCNP code in calculating same material form were validated and confirmed. Secondly, the neutron efficiency multiplication factors in the natural condition and accident condition were calculated and analyzed, and the safety in transport process was evaluated by taking conservative suppose of nuclear critical safety. The calculation results show that the max value of k eff for UX-30 transport freight package is less than the subcritical limiting value, and the UX-30 transport freight package is in the state of subcritical safety. Moreover, the critical safety index (CSI) for UX-30 package can define zero based on the definition of critical safety index. (authors)

  16. Boundary conditions for pathways, safety analysis and basic criteria for low-level radiation waste site selection

    International Nuclear Information System (INIS)

    Saverot, P.

    1994-01-01

    There are three successive periods in the life of a disposal facility: the operating period, the institutional control period and the unrestricted site access period. The purpose of safety analysis of the disposal facility is to ensure that the radiological impacts for each period in the life of the facility are acceptable under all circumstances. Founded on a deterministic approach, this analysis leads to a determination of the maximum quantity of each radionuclide present in the facility at the beginning of the institutional control period in order for the impacts to be considered acceptable. Safety analysis involves the calculation of the radiological impacts of a given radiological inventory under a selected scenario, from all plausible scenarios of radionuclide migration to the environment in both normal and accident conditions, and taking into account other specified variables. The calculation itself involves an assessment of the quantities of radionuclides that could be released to the environment under the specific scenario selected and following identified pathways, and a determination of the resultant exposure, both internal and external, to the public. An iterative approach is used in the performance of pathways analyses. If the pathways analyses result in unacceptable radiological impacts, either the radiological inventory of the site is reduced or barrier characteristics not previously factored into the analysis are taken into account. New pathways analyses are then performed until the results are within the acceptable range. Once accepted by the safety authorities, the radiological inventory becomes the radiological capacity, which is the approved quantities of specific radionuclides that may be disposed of at the site. The following elaborates on the boundary conditions used in safety analyses and describes the types of pathways analyses performed for a LLW disposal facility

  17. 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.

  18. 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)

  19. 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)

  20. 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.

  1. Implementing partnerships in nonreactor facility safety analyses

    International Nuclear Information System (INIS)

    Courtney, J.C.; Perry, W.H.; Phipps, R.D.

    1996-01-01

    Faculty and students from LSU have been participating in nuclear safety analyses and radiation protection projects at ANL-W at INEL since 1973. A mutually beneficial relationship has evolved that has resulted in generation of safety-related studies acceptable to Argonne and DOE, NRC, and state regulatory groups. Most of the safety projects have involved the Hot Fuel Examination Facility or the Fuel Conditioning Facility; both are hot cells that receive spent fuel from EBR-II. A table shows some of the major projects at ANL-W that involved LSU students and faculty

  2. 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

  3. 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

  4. Extended probabilistic system assessment calculations within the SKI project-90

    International Nuclear Information System (INIS)

    Pereira, A.

    1993-03-01

    The probabilistic system assessment calculation reported in the SKI Project-90 final documents were restricted to the following nuclides: 14 C, 129 I, 135 Cs, 237 Np and 240 Pu. In this report we have extended those calculations to another five nuclides: 79 Se, 243 Am, 240 Pu, 93 Zr and 99 Tc. The execution of probabilistic assessment calculations integrated in the context of SKIs first safety analysis exercise of an hypothetic final repository for high-level nuclear waste in Sweden, was a learning experience of relevance for the conduction of probabilistic safety assessment in future exercises. Some major conclusions and viewpoints of future need related with probabilistic assessment were withdrawn from this work and are presented in our report

  5. Safety Performance Indicator for alcohol in road accidents--international comparison, validity and data quality.

    Science.gov (United States)

    Assum, Terje; Sørensen, Michael

    2010-03-01

    Safety Performance Indicators, SPIs, are developed for various areas within road safety such as speed, car occupant protection, alcohol and drugs, vehicle safety, etc. SPIs can be used to indicate the road safety situation and to compare road safety performance between countries and over time and to understand the process leading to accidents, helping to select the measures to reduce them. This article describes an alcohol SPI defined as the percentage of fatalities resulting from accidents involving at least one driver impaired by alcohol. The calculation of the alcohol SPI for 26 European countries shows that the SPI varies from 4.4% in Bulgaria to 72.2% in Italy. These results raise the question if the results reflect the real situation or if there is a methodological explanation. To answer this question three different studies were carried out: comparison with other alcohol SPIs, in-depth studies of data quality in seven selected countries, and a study of correlations between the SPI and influencing factors. These studies indicate clearly that there is a need to improve quality of the data used for the alcohol SPI. Most importantly, the total number of drivers involved in fatal accidents, the number tested for alcohol and the number not tested, should be reported, in addition to the number of alcohol positive and negative drivers among those tested. Until these improvements are made, the validity of this SPI seems poor and comparison of the alcohol SPI results across countries should be made with caution. Copyright 2009 Elsevier Ltd. All rights reserved.

  6. Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues

    International Nuclear Information System (INIS)

    1992-12-01

    This report is to provide a comprehensive description of the implementation and verification status of Three Mile Island (TMI) Action Plan requirements, safety issues designated as Unresolved Safety Issues (USIs), Generic Safety Issues(GSIs), and other Multiplant Actions (MPAs) that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

  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. 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

  9. 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.

  10. 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

  11. Natural safety indicators and their application to repository safety cases

    International Nuclear Information System (INIS)

    Miller, B.

    2002-01-01

    Radiological dose and risk are the standard end-points calculated in all performance assessments. Their calculation requires, however, assumptions to be made for future human behaviour. To complement dose and risk, other safety indicators have been suggested which do not require such assumptions to be made. One proposed set of safety indicators are the concentrations and fluxes of naturally-occurring chemical species in the environment which may be compared with the performance assessment predictions of repository releases. Such comparisons can be valid because both the natural and repository species would occur in the same system and their transport behaviour would be controlled by exactly the same processes at the same rates. Although simple in concept, there is currently no consensus on the most appropriate comparisons to make or on the interpretation of such comparisons. A number of national and international research projects are evaluating this proposed approach, including an IAEA Co-ordinated Research Programme. These projects suggest that that the approach appears to be workable and that it may be a valuable component of a safety case, complementing the dose and risk presentations. Further work is, however, necessary to develop the approach to a level where it may be confidently applied in further performance assessments in a consistent and methodical manner. (author)

  12. 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

  13. Safety-I, Safety-II and Resilience Engineering.

    Science.gov (United States)

    Patterson, Mary; Deutsch, Ellen S

    2015-12-01

    In the quest to continually improve the health care delivered to patients, it is important to understand "what went wrong," also known as Safety-I, when there are undesired outcomes, but it is also important to understand, and optimize "what went right," also known as Safety-II. The difference between Safety-I and Safety-II are philosophical as well as pragmatic. Improving health care delivery involves understanding that health care delivery is a complex adaptive system; components of that system impact, and are impacted by, the actions of other components of the system. Challenges to optimal care include regular, irregular and unexampled threats. This article addresses the dangers of brittleness and miscalibration, as well as the value of adaptive capacity and margin. These qualities can, respectively, detract from or contribute to the emergence of organizational resilience. Resilience is characterized by the ability to monitor, react, anticipate, and learn. Finally, this article celebrates the importance of humans, who make use of system capabilities and proactively mitigate the effects of system limitations to contribute to successful outcomes. Copyright © 2015 Mosby, Inc. All rights reserved.

  14. 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

  15. Safety assessment for the above ground storage of Cadmium Safety and Control Rods at the Solid Waste Management Facility

    International Nuclear Information System (INIS)

    Shaw, K.W.

    1993-11-01

    The mission of the Savannah River Site is changing from radioisotope production to waste management and environmental restoration. As such, Reactor Engineering has recently developed a plan to transfer the safety and control rods from the C, K, L, and P reactor disassembly basin areas to the Transuranic (TRU) Waste Storage Pads for long-term, retrievable storage. The TRU pads are located within the Solid Waste Management Facilities at the Savannah River Site. An Unreviewed Safety Question (USQ) Safety Evaluation has been performed for the proposed disassembly basin operations phase of the Cadmium Safety and Control Rod Project. The USQ screening identified a required change to the authorization basis; however, the Proposed Activity does not involve a positive USQ Safety Evaluation. A Hazard Assessment for the Cadmium Safety and Control Rod Project determined that the above-ground storage of the cadmium rods results in no change in hazard level at the TRU pads. A Safety Assessment that specifically addresses the storage (at the TRU pads) phase of the Cadmium Safety and Control Rod Project has been performed. Results of the Safety Assessment support the conclusion that a positive USQ is not involved as a result of the Proposed Activity

  16. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  17. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  18. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  19. Calculation of Hazard Category 2/3 Threshold Quantities Using Contemporary Dosimetric Data

    Energy Technology Data Exchange (ETDEWEB)

    Walker, William C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-11-01

    The purpose of this report is to describe the methodology and selection of input data utilized to calculate updated Hazard Category 2 and Hazard Category 3 Threshold Quantities (TQs) using contemporary dosimetric information. The calculation of the updated TQs will be considered for use in the revision to the Department of Energy (DOE) Technical Standard (STD-) 1027-92 Change Notice (CN)-1, “Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports.” The updated TQs documented in this report complement an effort previously undertaken by the National Nuclear Security Administration (NNSA), which in 2014 issued revised Supplemental Guidance documenting the calculation of updated TQs for approximately 100 radionuclides listed in DOE-STD-1027-92, CN-1. The calculations documented in this report complement the NNSA effort by expanding the set of radionuclides to more than 1,250 radionuclides with a published TQ. The development of this report was sponsored by the Department of Energy’s Office of Nuclear Safety (AU-30) within the Associate Under Secretary for Environment, Health, Safety, and Security organization.

  20. Current activities and future trends in reliability analysis and probabilistic safety assessment in Hungary

    International Nuclear Information System (INIS)

    Hollo, E.; Toth, J.

    1986-01-01

    In Hungary reliability analysis (RA) and probabilistic safety assessment (PSA) of nuclear power plants was initiated 3 years ago. First, computer codes for automatic fault tree analysis (CAT, PREP) and numerical evaluation (REMO, KITT1,2) were adapted. Two main case studies - detailed availability/reliability calculation of diesel sets and analysis of safety systems influencing event sequences induced by large LOCA - were performed. Input failure data were taken from publications, a need for failure and reliability data bank was revealed. Current and future activities involves: setup of national data bank for WWER-440 units; full-scope level-I PSA of PAKS NPP in Hungary; operational safety assessment of particular problems at PAKS NPP. In the present article the state of RA and PSA activities in Hungary, as well as the main objectives of ongoing work are described. A need for international cooperation (for unified data collection of WWER-440 units) and for IAEA support (within Interregional Program INT/9/063) is emphasized. (author)

  1. What we now know (and don't know) about reactor safety

    International Nuclear Information System (INIS)

    Budnitz, R.J.

    1984-01-01

    This paper deals with the likelihood and consequences of major reactor accidents. Our understanding has advanced markedly in the past decade, largely because of the use of probabilistic risk assessment. There is now a good understanding of accident sequences, despite some important gaps. The insights achieved have in many ways revolutionized the reactor safety community's view of the relative importance of various safety issues. Understanding of severe accident phenomena is still inadequate, but experiments and development of analytical methods will soon enable us to calculate better the phenomena involved in core degradation, core melting, containment behavior, and the ultimate fate of fission products released from a melted core. Also we are now gaining important insights into analysis of human errors and external initiating events. Thus the next decade will see another quantum jump in our understanding, rivalling the progress of the last decade. (author)

  2. 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.

  3. Application of a general purpose user's version of the EGS4 code system to a photon skyshine benchmarking calculation

    International Nuclear Information System (INIS)

    Nojiri, I.; Fukasaku, Y.; Narita, O.

    1994-01-01

    A general purpose user's version of the EGS4 code system has been developed to make EGS4 easily applicable to the safety analysis of nuclear fuel cycle facilities. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with Kansas State University (KSU) photon skyshine experiment of 1977. The results of the simulation showed that this version of EGS4 would be appicable to the skyshine calculation. (author)

  4. Safety Review related to Commercial Grade Digital Equipment in Safety System

    International Nuclear Information System (INIS)

    Yu, Yeongjin; Park, Hyunshin; Yu, Yeongjin; Lee, Jaeheung

    2013-01-01

    The upgrades or replacement of I and C systems on safety system typically involve digital equipment developed in accordance with non-nuclear standards. However, the use of commercial grade digital equipment could include the vulnerability for software common-mode failure, electromagnetic interference and unanticipated problems. Although guidelines and standards for dedication methods of commercial grade digital equipment are provided, there are some difficulties to apply the methods to commercial grade digital equipment for safety system. This paper focuses on regulatory guidelines and relevant documents for commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. This paper focuses on KINS regulatory guides and relevant documents for dedication of commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. Dedication including critical characteristics is required to use the commercial grade digital equipment on safety system in accordance with KEPIC ENB 6370 and EPRI TR-106439. The dedication process should be controlled in a configuration management process. Appropriate methods, criteria and evaluation result should be provided to verify acceptability of the commercial digital equipment used for safety function

  5. 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

  6. Electronics and data processing for safety

    International Nuclear Information System (INIS)

    1995-01-01

    Industrial installations, and in particular installations involving risk, are more and more monitored and controlled by computerized systems. The use of such systems raises questions about their contribution to the installation safety and about the qualities required in these systems to avoid additional risk. The February 1995 Electronics Days were organized by the CEA-LETI Department of Electronics and Nuclear Instrumentation to try to answer these questions. Four sessions were organized on the following topics: computerized systems and functioning safety, components and architectures, softwares and norms, and tools and methods. Only the communications dealing with the safety of computerized systems and components involved in nuclear applications have been retained (17 over 36). (J.S.)

  7. Generic radiation safety design for SSRL synchrotron radiation beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Liu, James C. [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States)]. E-mail: james@slac.stanford.edu; Fasso, Alberto [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Khater, Hesham [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Prinz, Alyssa [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Rokni, Sayed [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States)

    2006-12-15

    To allow for a conservative, simple, uniform, consistent, efficient radiation safety design for all SSRL beamlines, a generic approach has been developed, considering both synchrotron radiation (SR) and gas bremsstrahlung (GB) hazards. To develop the methodology and rules needed for generic beamline design, analytic models, the STAC8 code, and the FLUKA Monte Carlo code were used to pre-calculate sets of curves and tables that can be looked up for each beamline safety design. Conservative beam parameters and standard targets and geometries were used in the calculations. This paper presents the SPEAR3 beamline parameters that were considered in the design, the safety design considerations, and the main pre-calculated results that are needed for generic shielding design. In the end, the rules and practices for generic SSRL beamline design are summarized.

  8. 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.

  9. 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)

  10. DEVELOPING SAFETY INDICATORS ON THE BASIS OF THE ICAO RECOMMENDATIONS

    Directory of Open Access Journals (Sweden)

    V. D. Sharov

    2014-01-01

    Full Text Available The article offers direct use of the recommendations of SMM ICAO Doc.9859, 3rd ed. 2013, for calculation the target and alert levels of safety indicators. Examples of calculation based on data of 2011 and monitoring of the current indicators during 2012 are presented. Safety indicators for airlines in terms of “numbers of incidents per 1000 flight hours” could be calculated on the basis of the state values through the «coefficient of conformity».

  11. Criticality Safety in the Handling of Fissile Material. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-05-15

    This Safety Guide provides guidance and recommendations on how to meet the relevant requirements for ensuring subcriticality when dealing with fissile material and for planning the response to criticality accidents. The guidance and recommendations are applicable to both regulatory bodies and operating organizations. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences of this if it were to occur. The Safety Guide makes recommendations on how to ensure subcriticality in systems involving fissile materials during normal operation, anticipated operational occurrences, and, in the case of accident conditions, within design basis accidents, from initial design through commissioning, operation, and decommissioning and disposal.

  12. Nuclear power plant's safety and risk (requirements of safety and reliability)

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1977-01-01

    Starting out from the given safety objectives as they have evolved during the past few years and from the present legal and regulatory provisions for the construction and operation of nuclear power plants, the hazards involved in regular operation, accidents and emergency situations are discussed. In compliance with the positive safety balance of nuclear power plants in the FRG, special attention is focused on the preventive safety analysis within the frame of the nuclear licensing procedure. Reference is made to the beginnings of a comprehensive hazard concept for an unbiased plant assessment. Emergency situations are discussed from the point of view of general hazard comparisons. (orig.) [de

  13. 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.

  14. 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...

  15. Development and applications of a safety assessment system for promoting safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Takano, Ken-ichi; Hasegawa, Naoko; Hirose, Ayako; Hayase, Ken-ichi

    2004-01-01

    For past five years, CRIEPI has been continuing efforts to develop and make applications of a 'safety assessment system' which enable to measure the safety level of organization. This report describe about frame of the system, assessment results and its reliability, and relation between labor accident rate in the site and total safety index (TSI), which can be obtained by the principal factors analysis. The safety assessment in this report is based on questionnaire survey of employee. The format and concrete questionnaires were developed using existing literatures including organizational assessment tools. The tailored questionnaire format involved 124 questionnaire items. The assessment results could be considered as a well indicator of the safety level of organization, safety management, and safety awareness of employee. (author)

  16. 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

  17. 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)

  18. 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.

  19. 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...

  20. A simple reliability block diagram method for safety integrity verification

    International Nuclear Information System (INIS)

    Guo Haitao; Yang Xianhui

    2007-01-01

    IEC 61508 requires safety integrity verification for safety related systems to be a necessary procedure in safety life cycle. PFD avg must be calculated to verify the safety integrity level (SIL). Since IEC 61508-6 does not give detailed explanations of the definitions and PFD avg calculations for its examples, it is difficult for common reliability or safety engineers to understand when they use the standard as guidance in practice. A method using reliability block diagram is investigated in this study in order to provide a clear and feasible way of PFD avg calculation and help those who take IEC 61508-6 as their guidance. The method finds mean down times (MDTs) of both channel and voted group first and then PFD avg . The calculated results of various voted groups are compared with those in IEC61508 part 6 and Ref. [Zhang T, Long W, Sato Y. Availability of systems with self-diagnostic components-applying Markov model to IEC 61508-6. Reliab Eng System Saf 2003;80(2):133-41]. An interesting outcome can be realized from the comparison. Furthermore, although differences in MDT of voted groups exist between IEC 61508-6 and this paper, PFD avg of voted groups are comparatively close. With detailed description, the method of RBD presented can be applied to the quantitative SIL verification, showing a similarity of the method in IEC 61508-6

  1. 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.)

  2. 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.)

  3. Safety analysis in support of regulatory decision marking

    International Nuclear Information System (INIS)

    Pomier Baez, L.; Troncoso Fleitas, M.; Valhuerdi Debesa, C.; Valle Cepero, R.; Hernandez, J.L.

    1996-01-01

    Features of different safety analysis techniques by means of calculation thermohydraulic a probabilistic and severe accidents used in the safety assessment, as well as the development of these techniques in Cuba and their use in support of regulatory decision making are presented

  4. Knowledge representation in safety assessment: improving transparency and traceability

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  5. Knowledge representation in safety assessment: improving transparency and traceability

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Measurement and analysis of CEFR safety and shim rod worth

    International Nuclear Information System (INIS)

    Chen Yiyu; Yang Yong; Gang Zhi; Xu Li; Yang Xiaoyan; Zhou Keyuan; Hu Dingsheng

    2013-01-01

    The reactivity worth of safety rods and shim rods in critical phase and operating phase was calculated respectively using Monte Carlo program in this paper. In addition, the reactivity worth of safety rods and shim rods was measured by the rod drop-off method and period method. The experimental results are in good agreement with the calculated values with less than 5% error. It illustrates the high calculation precision of Monte Carlo program, which provides a practical reference for subsequent application of Monte Carlo program in future demonstration fast reactors. (authors)

  7. 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

  8. CFD Analysis of the Safety Injection Tank and Fluidic Device

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Oan; Nietiadi, Yohanes Setiawan; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Addad, Yacine [KUSTAR, Abu Dhabi (United Arab Emirates)

    2016-05-15

    One of the most important components in the ECCS is the safety injection tank (SIT). Inside the SIT, a fluidic device is installed, which passively controls the mass flow of the safety injection and eliminates the need for low pressure safety injection pumps. As more passive safety mechanisms are being pursued, it has become more important to understand flow structure and the loss mechanism within the fluidic device. Current computational fluid dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study proposes to find a more exact result using CFD and more realistic modeling to predict the performance during accident scenarios more accurately. The safety injection tank with fluidic device was analyzed thoroughly using CFD. The preliminary calculation used 60,000 meshes for the initial test calculation. The results fit the experimental results surprisingly despite its coarse grid. Nonetheless, the mesh resolution was increased to capture the vortex in the fluidic device precisely. Once a detailed CFD computation is finished, a small-scale experiment will be conducted for the given conditions. Using the experimental results and the CFD model, physical models can be improved to fit the results more accurately.

  9. Safety analysis of accident localization system

    International Nuclear Information System (INIS)

    1999-01-01

    A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

  10. Preliminary safety analysis of the HTTR-IS nuclear hydrogen production system

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Tachibana, Yukio; Sakaba, Nariaki

    2010-06-01

    Japan Atomic Energy Agency is planning to demonstrate hydrogen production by thermochemical water-splitting IS process utilizing heat from the high-temperature gas-cooled reactor HTTR (HTTR-IS system). The previous study identified that the HTTR modification due to the coupling of hydrogen production plant requires an additional safety review since the scenario and quantitative values of the evaluation items would be altered from the original HTTR safety review. Hence, preliminary safety analyses are conducted by using the system analysis code. Calculation results showed that evaluation items such as a coolant pressure, temperatures of heat transfer tubes at the pressure boundary, etc., did not exceed allowable values. Also, the peak fuel temperature did not exceed allowable value and therefore the reactor core was not damaged and cooled sufficiently. This report compiles calculation conditions, event scenarios and the calculation results of the preliminary safety analysis. (author)

  11. 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.

  12. Preliminary investigation on reliability assessment of passive safety system

    International Nuclear Information System (INIS)

    Huang Changfan; Kuang Bo

    2012-01-01

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

  13. Safety analysis methodology with assessment of the impact of the prediction errors of relevant parameters

    International Nuclear Information System (INIS)

    Galia, A.V.

    2011-01-01

    The best estimate plus uncertainty approach (BEAU) requires the use of extensive resources and therefore it is usually applied for cases in which the available safety margin obtained with a conservative methodology can be questioned. Outside the BEAU methodology, there is not a clear approach on how to deal with the issue of considering the uncertainties resulting from prediction errors in the safety analyses performed for licensing submissions. However, the regulatory document RD-310 mentions that the analysis method shall account for uncertainties in the analysis data and models. A possible approach is presented, that is simple and reasonable, representing just the author's views, to take into account the impact of prediction errors and other uncertainties when performing safety analysis in line with regulatory requirements. The approach proposes taking into account the prediction error of relevant parameters. Relevant parameters would be those plant parameters that are surveyed and are used to initiate the action of a mitigating system or those that are representative of the most challenging phenomena for the integrity of a fission barrier. Examples of the application of the methodology are presented involving a comparison between the results with the new approach and a best estimate calculation during the blowdown phase for two small breaks in a generic CANDU 6 station. The calculations are performed with the CATHENA computer code. (author)

  14. Assessment of safety culture at INPP

    International Nuclear Information System (INIS)

    Lesin, S.

    2002-01-01

    Safety Culture covers all main directions of plant activities and the plant departments involved through integration into the INPP Quality Assurance System. Safety Culture is represented by three components. The first is the clear INPP Safety and Quality Assurance Policy. Based on the Policy INPP is safely operated and managers' actions firstly aim at safety assurance. The second component is based on personal responsibility for safety and attitude of each employee of the plant. The third component is based on commitment to safety and competence of managers and employees of the plant. This component links the first two to ensure efficient management of safety at the plant. The above mentioned components including the elements which may significantly affect Safety Culture are also presented in the attachment. The concept of such model implies understanding of effect of different factors on the level of Safety Culture in the organization. In order to continuously correct safety problems, self-assessment of the Safety Culture level is performed at regular intervals. (author)

  15. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition); Bezopasnost' atomnykh ehlektrostantsij: proektirovanie. Konkretnye trebovaniya bezopasnosti

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  16. An analysis of electronic health record-related patient safety concerns

    Science.gov (United States)

    Meeks, Derek W; Smith, Michael W; Taylor, Lesley; Sittig, Dean F; Scott, Jean M; Singh, Hardeep

    2014-01-01

    Objective A recent Institute of Medicine report called for attention to safety issues related to electronic health records (EHRs). We analyzed EHR-related safety concerns reported within a large, integrated healthcare system. Methods The Informatics Patient Safety Office of the Veterans Health Administration (VA) maintains a non-punitive, voluntary reporting system to collect and investigate EHR-related safety concerns (ie, adverse events, potential events, and near misses). We analyzed completed investigations using an eight-dimension sociotechnical conceptual model that accounted for both technical and non-technical dimensions of safety. Using the framework analysis approach to qualitative data, we identified emergent and recurring safety concerns common to multiple reports. Results We extracted 100 consecutive, unique, closed investigations between August 2009 and May 2013 from 344 reported incidents. Seventy-four involved unsafe technology and 25 involved unsafe use of technology. A majority (70%) involved two or more model dimensions. Most often, non-technical dimensions such as workflow, policies, and personnel interacted in a complex fashion with technical dimensions such as software/hardware, content, and user interface to produce safety concerns. Most (94%) safety concerns related to either unmet data-display needs in the EHR (ie, displayed information available to the end user failed to reduce uncertainty or led to increased potential for patient harm), software upgrades or modifications, data transmission between components of the EHR, or ‘hidden dependencies’ within the EHR. Discussion EHR-related safety concerns involving both unsafe technology and unsafe use of technology persist long after ‘go-live’ and despite the sophisticated EHR infrastructure represented in our data source. Currently, few healthcare institutions have reporting and analysis capabilities similar to the VA. Conclusions Because EHR-related safety concerns have complex

  17. Calculation of anti-seismic design for Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    Li Shuian

    2002-01-01

    The author describes the reactor safety rule, safety regulation and design code that must be observed to anti-seismic design in Xi'an pulsed reactor. It includes the classification of reactor installation, determination of seismic loads, calculate contents, program, method, results and synthetically evaluation. According to the different anti-seismic structure character of reactor installation, an appropriate method was selected to calculate the seismic response. The results were evaluated synthetically using the design code and design requirement. The evaluate results showed that the anti-seismic design function of reactor installation of Xi'an pules reactor is well, and the structure integrality and normal property of reactor installation can be protect under the designed classification of the earthquake

  18. 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

  19. 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

  20. Nuclear safety guide TID-7016 Revision 2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1980-01-01

    The present revision of TID-7016 Nuclear Safety Guide is discussed. This Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams, experienced in the field. The reader will find a significant change in the character of information presented in this version. Nuclear Criticality Safety has matured in the past twelve years. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available, information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the former Guides by employing appropriate safety factors. In fact, it becomes incumbent on the Criticality Safety Specialist to necessarily impose safety factors consistent with the possible normal and abnormal credible contingencies of an operation as revealed by his evaluation. In its present form the Guide easily becomes a suitable module in any compendium or handbook tailored for internal use by organizations. It is hoped the Guide will continue to serve immediate needs and will encourage continuing and more comprehensive efforts toward organizing nuclear criticality safety information

  1. The role of psychological factors in workplace safety.

    Science.gov (United States)

    Kotzé, Martina; Steyn, Leon

    2013-01-01

    Workplace safety researchers and practitioners generally agree that it is necessary to understand the psychological factors that influence people's workplace safety behaviour. Yet, the search for reliable individual differences regarding psychological factors associated with workplace safety has lead to sparse results and inconclusive findings. The aim of this study was to investigate whether there are differences between the psychological factors, cognitive ability, personality and work-wellness of employees involved in workplace incidents and accidents and/or driver vehicle accidents and those who are not. The study population (N = 279) consisted of employees employed at an electricity supply organisation in South Africa. Mann-Whitney U-test and one-way ANOVA were conducted to determine the differences in the respective psychological factors between the groups. These results showed that cognitive ability did not seem to play a role in workplace incident/accident involvement, including driver vehicle accidents, while the wellness factors burnout and sense of coherence, as well as certain personality traits, namely conscientiousness, pragmatic and gregariousness play a statistically significant role in individuals' involvement in workplace incidents/accidents/driver vehicle accidents. Safety practitioners, managers and human resource specialists should take cognisance of the role of specifically work-wellness in workplace safety behaviour, as management can influence these negative states that are often caused by continuously stressful situations, and subsequently enhance work place safety.

  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. Comparison of the safety information on drug labels in three developed countries: The USA, UK and Canada

    Directory of Open Access Journals (Sweden)

    Thamir M. Alshammari

    2017-12-01

    Full Text Available The safety information on drug labels of a company marketing the same drugs in different countries is sometimes different. The aim of the present study is to understand the differences in the volume and content of safety information on the drug labels from the same manufacturers in three developed countries: the United States of America (USA, the United Kingdom (UK and Canada. This study involved the calculation of the proportion of total safety information (PSI and of contraindications (PCI in comparison to all information on the label and the percentage of boxed warnings (PBW among the 100 labels studied from each country. The PSI on the labels of different countries is different with USA labels bearing lesser value PSI and UK labels bearing higher value PSI. The qualitative information provided on these drug labels from each country in ‘contraindications’ sections, ‘boxed/serious warnings’ and ‘overdosage’ sections presented differences in the information provided on most of the labels. We have found distinct differences between the safety information available on drug labels in terms of volume and content. We conclude that the safety information for the same products should be standardised across all countries.

  4. [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

  5. Evaluation of operating experience with safety values

    International Nuclear Information System (INIS)

    Bung, W.; Hoemke, P.; Oberender, W.; Paul, H.; Rueter, W.

    1985-01-01

    This report describes statistical investigations of 2076 functional tests carried out on power operated safety valves in conventional power plants in 1972 until 1983 with special regard to Common Mode-Failures. The results clearly show that Common Mode-Failures play an important part of non-availability for the controlled safety valves, especially in the control system. The 'Deutsche Risikostudie' does not consider any Common Mode-Failures of the primary safety valves. However there is no significant increase of the risk resulted by the primary safety valves in the 'Referenzanlage' if the calculated Common Mode-Failures probabilities are considered. (orig.) [de

  6. 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)

  7. 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.

  8. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  9. Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)

    2012-07-01

    In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

  10. Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.

    2012-01-01

    In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

  11. 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

  12. Risk communication activities toward nuclear safety in Tokai: your safety is our safety

    International Nuclear Information System (INIS)

    Tsuchiya, T.

    2007-01-01

    As several decades have passed since the construction of nuclear power plants began, residents have become gradually less interested in nuclear safety. The Tokai criticality accident in 1909, however, had roused residents in Tokai-Mura to realize that they live with nuclear technology risks. To prepare a field of risk communication, the Tokai-Mura C 3 project began as a pilot research project supported by NISA. Alter the project ended, we are continuing risk. communication activities as a non-profit organisation. The most important activity of C 3 project is the citizen's inspection programme for nuclear related facilities. This programme was decided by participants who voluntarily applied to the project. The concept of the citizen's inspection programme is 'not the usual facility tours'. Participants are involved from the planning stage and continue to communicate with workers of the inspected nuclear facility. Since 2003, we have conducted six programmes for five nuclear related organisations. Participants evaluated that radiation protection measures were near good but there were some problems concerning the worker's safety and safety culture, and proposed a mixture of advice based on personal experience. Some advice was accepted and it did improve the facility's safety measures. Other suggestions were not agreed upon by nuclear organisations. The reason lies in the difference of concept between the nuclear expert's 'safety' and the citizen's 'safety'. Residents do not worry about radiation only, but also about the facility's safety as a whole including the worker's safety. They say, 'If the workers are not safe, you also are unable to protect us'. Although the disagreement remained, the participants and the nuclear industry learned much about each other. Participating citizens received a substantial amount of knowledge about the nuclear industry and its safety measures, and feel the credibility and openness of the nuclear industry. On the other hand, the nuclear

  13. 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,

  14. Preliminary safety analysis of unscrammed events for KLFR

    International Nuclear Information System (INIS)

    Kim, S.J.; Ha, G.S.

    2005-01-01

    The report presents the design features of KLFR; Safety Analysis Code; steady-state calculation results and analysis results of unscrammed events. The calculations of the steady-state and unscrammed events have been performed for the conceptual design of KLFR using SSC-K code. UTOP event results in no fuel damage and no centre-line melting. The inherent safety features are demonstrated through the analysis of ULOHS event. Although the analysis of ULOF has much uncertainties in the pump design, the analysis results show the inherent safety characteristics. 6% flow of rated flow of natural circulation is formed in the case of ULOF. In the metallic fuel rod, the cladding temperature is somewhat high due to the low heat transfer coefficient of lead. ULOHS event should be considered in design of RVACS for long-term cooling

  15. 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

  16. Fusion Safety Program annual report, fiscal year 1992

    International Nuclear Information System (INIS)

    Holland, D.F.; Cadwallader, L.C.; Herring, J.S.; Longhurst, G.R.; McCarthy, K.A.; Merrill, B.J.; Piet, S.J.

    1993-01-01

    This report summarizes the major activities of the Fusion Safety Program in fiscal year 1992. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and EG ampersand G Idaho, Inc. is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL and in participating organizations including the Westinghouse Hanford Company at the Hanford Engineering Development Laboratory, the Massachusetts Institute of Technology, and the University of Wisconsin. The technical areas covered in the report include tritium safety, activation product release, reactions involving beryllium, reactions involving lithium breeding materials, safety of fusion magnet systems, plasma disruptions, risk assessment failure rate data base, and computer code development for reactor transients. Also included in the report is a summary of the safety and environmental studies performed by the INEL for the Tokamak Physics Experiments and the Tokamak Fusion Test Reactor, the safety analysis for the International Thermonuclear Experimental Reactor design, and the technical support for the ARIES commercial reactor design study

  17. Sn transport calculations on vector and parallel processors

    International Nuclear Information System (INIS)

    Rhoades, W.A.; Childs, R.L.

    1987-01-01

    The transport of radiation from the source to the location of people or equipment gives rise to some of the most challenging of calculations. A problem may involve as many as a billion unknowns, each evaluated several times to resolve interdependence. Such calculations run many hours on a Cray computer, and a typical study involves many such calculations. This paper will discuss the steps taken to vectorize the DOT code, which solves transport problems in two space dimensions (2-D); the extension of this code to 3-D; and the plans for extension to parallel processors

  18. Best-estimated multi-dimensional calculation during LB LOCA for APR1400

    International Nuclear Information System (INIS)

    Oh, D. Y.; Bang, Y. S.; Cheong, A. J.; Woong, S.; Korea, W.

    2010-01-01

    Best-estimated (BE) calculation with uncertainty quantification for the emergency core cooling system (ECCS) performance analysis during Loss of Coolant Accident (LOCA) is more broadly used in nuclear industries and regulations. In Korea, demand on regulatory audit calculation is continuously increasing to support the safety review for life extension, power up-rating and advanced nuclear reactor design. The thermal-hydraulic system code, MARS (Multi-dimensional Analysis of Reactor Safety), with multi-dimensional capability is used for audit calculation. It achieves to describe the complicated phenomena in reactor coolant system by very effectively consolidating the one dimensional RELAP5/MOD3 with the multidimensional COBRA-TF codes. The advanced power reactors (APR1400) to be evaluated has four separated hydraulic trains of the high pressure injection system (HPSI) with direct vessel injection (DVI) which is different from the existing commercial PWRs. Also, the therma-hydraulic behavior of DVI plant would be considerably different from that of a cold-leg safety injection since the low pressure safety injection system are eliminated and the high pressure safety flow are injected into the specific elevation of reactor vessel downcomer. The ECCS bypass induced by the downcomer boiling due to hot wall heating of reactor vessel during reflooding phase is one of the important phenomena which should be considered in DVI plants. Therefore, in this study, BE calculation with one-dimensional (1-D) and multi-dimensional (multi-D) MARS models during LBLOCA are performed for APR1400 plant. In the multi-D evaluation, the reactor vessel is modeled by multi-D components and the specific treatment of flow path inside reactor vessel, e.g., upper guide structure, is essential. The concept of hot zone is adopted to simulate the limiting thermal-hydraulic conditions surrounding hot rod, which is similar to hot channel in 1-D. Also, alternative treatment of the hot rods in multi-D is

  19. Methods for tornado frequency calculation of nuclear power plant

    International Nuclear Information System (INIS)

    Liu Haibin; Li Lin

    2012-01-01

    In order to take probabilistic safety assessment of nuclear power plant tornado attack event, a method to calculate tornado frequency of nuclear power plant is introduced based on HAD 101/10 and NUREG/CR-4839 references. This method can consider history tornado frequency of the plant area, construction dimension, intensity various along with tornado path and area distribution and so on and calculate the frequency of different scale tornado. (authors)

  20. USSR orders computers to improve nuclear safety

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Control Data Corp (CDC) has received an order valued at $32-million from the Soviet Union for six Cyber 962 mainframe computer systems to be used to increase the safety of civilian nuclear powerplants. The firm is now waiting for approval of the contract by the US government and Western Allies. The computers, ordered by the Soviet Research and Development Institute of Power Engineering (RDIPE), will analyze safety factors in the operation of nuclear reactors over a wide range of conditions. The Soviet Union's civilian nuclear program is one of the largest in the world, with over 50 plants in operation. Types of safety analyses the computers perform include: neutron-physics calculations, radiation-protection studies, stress analysis, reliability analysis of equipment and systems, ecological-impact calculations, transient analysis, and support activities for emergency response. They also include a simulator with realistic mathematical models of Soviet nuclear powerplants to improve operator training

  1. 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.

  2. Nuclear reaction models - source term estimation for safety design in accelerators

    International Nuclear Information System (INIS)

    Nandy, Maitreyee

    2013-01-01

    Accelerator driven subcritical system (ADSS) employs proton induced spallation reaction at a few GeV. Safety design of these systems involves source term estimation in two steps - multiple fragmentation of the target and n+γ emission through a fast process followed by statistical decay of the primary fragments. The prompt radiation field is estimated in the framework of quantum molecular dynamics (QMD) theory, intra-nuclear cascade or Monte Carlo calculations. A few nuclear reaction model codes used for this purpose are QMD, JQMD, Bertini, INCL4, PHITS, followed by statistical decay codes like ABLA, GEM, GEMINI, etc. In the case of electron accelerators photons and photoneutrons dominate the prompt radiation field. High energy photon yield through Bremsstrahlung is estimated in the framework of Born approximation while photoneutron production is calculated using giant dipole resonance and quasi-deuteron formation cross section. In this talk hybrid and exciton PEQ models and QMD formalism will be discussed briefly

  3. Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

    International Nuclear Information System (INIS)

    Lajtman, S.; Feretic, D.; Debrecin, N.

    1989-01-01

    This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

  4. 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

  5. [B-BS and occupational health and safety management systems].

    Science.gov (United States)

    Bacchetta, Adriano Paolo

    2010-01-01

    The objective of a SGSL is the "prevention" agreement as approach of "pro-active" toward the safety at work through the construction of an integrated managerial system in synergic an dynamic way with the business organization, according to continuous improvement principles. Nevertheless the adoption of a SGSL, not could guarantee by itself the obtainment of the full effectiveness than projected and every individual's adhesion to it, must guarantee it's personal involvement in proactive way, so that to succeed to actual really how much hypothesized to systemic level to increase the safety in firm. The objective of a behavioral safety process that comes to be integrated in a SGSL, it has the purpose to succeed in implementing in firm a process of cultural change that raises the workers social group fundamental safety value, producing an ample and full involvement of all in the activities of safety at work development. SGSL = Occupational Health and Safety Management System.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  7. 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

  8. TH-C-18C-01: MRI Safety

    Energy Technology Data Exchange (ETDEWEB)

    Pooley, R [Mayo Clinic, Jacksonville, FL (United States); Bernstein, M; Shu, Y; Gorny, K; Felmlee, J [Mayo Clinic, Rochester, MN (United States); Panda, A [Mayo Clinic, Arizona, Scottsdale, AZ (United States)

    2014-06-15

    Clinical diagnostic medical physicists may be responsible for implementing and maintaining a comprehensive MR safety program. Accrediting bodies including the ACR, IAC, Radsite and The Joint Commission each include aspects of MR Safety into their imaging accreditation programs; MIPPA regulations further raise the significance of non-compliance. In addition, The Joint Commission recently announced New and Revised Diagnostic Imaging Standards for accredited health care organizations which include aspects of MR Safety. Hospitals and clinics look to the physicist to understand guidelines, regulations and accreditation requirements related to MR safety. The clinical medical physicist plays a significant role in a clinical practice by understanding the physical basis for the risks and acting as a facilitator to successfully implement a safety program that provides well-planned siting, allows for the safe scanning of certain implanted devices, and helps radiologists manage specific patient exams. The MRI scanning of specific devices will be discussed including cardiac pacemakers and neurostimulators such as deep brain stimulators. Furthermore for sites involved in MR guided interventional procedures, the MR physicist plays an essential role to establish safe practices. Creating a framework for a safe MRI practice includes the review of actual safety incidents or close calls to determine methods for prevention in the future. Learning Objectives: Understand the requirements and recommendations related to MR safety from accrediting bodies and federal regulations. Understand the Medical Physicist's roles to ensure MR Safety. Identify best practices for dealing with implanted devices, including pacemakers and deep brain stimulators. Review aspects of MR safety involved in an MR guided interventional environment. Understand the important MR safety aspects in actual safety incidents or near misses.

  9. School Violence: The Role of Parental and Community Involvement

    Science.gov (United States)

    Lesneskie, Eric; Block, Steven

    2017-01-01

    This study utilizes the School Survey on Crime and Safety to identify variables that predict lower levels of violence from four domains: school security, school climate, parental involvement, and community involvement. Negative binomial regression was performed and the findings indicate that statistically significant results come from all four…

  10. Nuclear Safety Co-Ordination within Oak Ridge Operations Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W. A.; Pryor, W. A. [Research and Development Division, United States Atomic Energy Commission, Oak Ridge, TN (United States)

    1966-05-15

    The Oak Ridge Operations Office of the USAEC has within its jurisdiction multiple contractors and facilities for research and for the production of fissile materials for the atomic energy programme. Among these facilities are gaseous diffusion plants for the production of {sup 235}U-enriched uranium hexafluoride, plants for the fabrication of special components and fuel for research and production reactors, and laboratories for pilot plant studies and basic research in nuclear technology. One research laboratory is also actively engaged in criticality experimental programmes and has been a major contributor of criticality data for safety applications. These diversified programmes include the processing, fabrication and transport of practically all forms and isotopic enrichments of uranium in quantities commensurate with both laboratory and volume production requirements. Consequently, adequate nuclear safety control with reasonable economy for operations of this magnitude demands not only co-ordination and liaison between contractor and USAEC staffs, but a continuing reappraisal of safety applications in light of the most advanced information. This report outlines the role of the Oak Ridge Operations Office in these pursuits and describes as examples some specific problems in which this office co-ordinated actions necessary for their resolution. Other examples are given of parametric and procedural applications in plant processes and fissile shipments emphasizing the use of recent experimental or calculated data. These examples involve the use of mass and geometric variables, neutron absorbers and moderation control. Departures from limits specified in existing nuclear safety guides are made to advantage in light of new data, special equipment design, contingencies and acceptable risks. (author)

  11. Radiation shielding and safety design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Ouk; Gil, C. S.; Cho, Y. S.; Kim, D. H.; Kim, H. I.; Kim, J. W.; Lee, C. W.; Kim, K. Y.; Kim, B. H. [KAERI, Daejeon (Korea, Republic of)

    2011-07-15

    A benchmarking for the test facility, evaluations of the prompt radiation fields, evaluation of the induced activities in the facility, and estimation of the radiological impact on the environment were performed in this study. and the radiation safety analysis report for nuclear licensing was written based on this study. In the benchmark calculation, the neutron spectra was measured in the 20 Mev test facility and the measurements were compared with the computational results to verify the calculation system. In the evaluation of the prompt radiation fields, the shielding design for 100 MeV target rooms, evaluations of the leakage doses from the accidents and skyshine analysis were performed. The evaluation of the induced activities were performed for the coolant, inside air, structural materials, soil and ground-water. At last, the radiation safety analysis report was written based on results from these studies

  12. Roadside design in The Netherlands for enhancing safety : contribution to the conference `Traffic safety on Two Continents', Lisbon, Portugal, September 22-24, 1997.

    NARCIS (Netherlands)

    Schoon, C.C.

    1998-01-01

    Safety barriers are often used on motorways. Accident figures, however, show that a safety barrier is involved in approximately 20% of all fatal accidents. This paper considers safety barriers within the context of safe designs for shoulders on motorways. This research is related to the European

  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. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Shibata, Toshikazu.

    1982-01-01

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  15. 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.)

  16. 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

  17. Improvements in practical applicability of NSHEX: nodal transport calculation code for three-dimensional hexagonal-Z geometry

    International Nuclear Information System (INIS)

    Sugino, Kazuteru

    1998-07-01

    As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)

  18. Evaluating software for safety systems in nuclear power plants

    International Nuclear Information System (INIS)

    Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.

    1994-01-01

    In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper

  19. 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

  20. 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)

  1. 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...

  2. Safety issues of nuclear production of hydrogen

    International Nuclear Information System (INIS)

    Piera, Mireia; Martinez-Val, Jose M.; Jose Montes, Ma

    2006-01-01

    Hydrogen is not an uncommon issue in Nuclear Safety analysis, particularly in relation to severe accidents. On the other hand, hydrogen is a household name in the chemical industry, particularly in oil refineries, and is also a well known chemical element currently produced by steam reforming of natural gas, and other methods (such as coal gasification). In the not-too-distant future, hydrogen will have to be produced (by chemical reduction of water) using renewable and nuclear energy sources. In particular, nuclear fission seems to offer the cheapest way to provide the primary energy in the medium-term. Safety principles are fundamental guidelines in the design, construction and operation both of hydrogen facilities and nuclear power plants. When these two technologies are integrated, a complete safety analysis must consider not only the safety practices of each industry, but any interaction that could be established between them. In particular, any accident involving a sudden energy release from one of the facilities can affect the other. Release of dangerous substances (chemicals, radiotoxic effluents) can also pose safety problems. Although nuclear-produced hydrogen facilities will need specific approaches and detailed analysis on their safety features, a preliminary approach is presented in this paper. No significant roadblocks are identified that could hamper the deployment of this new industry, but some of the hydrogen production methods will involve very demanding safety standards

  3. Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

    2001-08-15

    Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

  4. Safety culture and subcontractor network governance in a complex safety critical project

    International Nuclear Information System (INIS)

    Oedewald, Pia; Gotcheva, Nadezhda

    2015-01-01

    In safety critical industries many activities are currently carried out by subcontractor networks. Nevertheless, there are few studies where the core dimensions of resilience would have been studied in safety critical network activities. This paper claims that engineering resilience into a system is largely about steering the development of culture of the system towards better ability to anticipate, monitor, respond and learn. Thus, safety culture literature has relevance in resilience engineering field. This paper analyzes practical and theoretical challenges in applying the concept of safety culture in a complex, dynamic network of subcontractors involved in the construction of a new nuclear power plant in Finland, Olkiluoto 3. The concept of safety culture is in focus since it is widely used in nuclear industry and bridges the scientific and practical interests. This paper approaches subcontractor networks as complex systems. However, the management model of the Olkiluoto 3 project is to a large degree a traditional top-down hierarchy, which creates a mismatch between the management approach and the characteristics of the system to be managed. New insights were drawn from network governance studies. - Highlights: • We studied a relevant topical subject safety culture in nuclear new build project. • We integrated safety science challenges and network governance studies. • We produced practicable insights in managing safety of subcontractor networks

  5. 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

  6. How good is BNFL's health and safety?

    International Nuclear Information System (INIS)

    Berry, R.J.; Coulston, D.J.

    1989-01-01

    A historical perspective of the development of health and safety philosophy and practice within the nuclear industry is followed by a description of the Company's safety management system. This involves three elements, policy, implementation and audit. Data given on discharges to the environment and occupational dose. (U.K.)

  7. Main factors affecting strong ground motion calculations: Critical review and assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadioun, B [DAS/SASC (France); Pecker, A [Societe Geodynamique et Structure (France)

    1990-07-01

    In the interests of guarding lives and property against the effects of earthquakes, building codes are frequently enforced when erecting conventional structures, usually calling for simple, static calculations. Where more vulnerable facilities are involved, the failure of which, or of parts of which, could subject the environment to harmful substances, more sophisticated methods are used to compute or verify their design, often accompanied by safety margins intended to compensate for uncertainties encountered at various stages of the analysis that begins with input seismic data and culminates with an effective anti-seismic design. The forthcoming discussion will deal with what is known of the characteristics of strong ground motion, highly variable according to context, without entering into the problems raised by seismotectonic studies, which actually constitute the first aspect that must be addressed when performing such an analysis. Our conclusion will be devoted to cogent R and D work in this area.

  8. Main factors affecting strong ground motion calculations: Critical review and assessment

    International Nuclear Information System (INIS)

    Mohammadioun, B.; Pecker, A.

    1990-01-01

    In the interests of guarding lives and property against the effects of earthquakes, building codes are frequently enforced when erecting conventional structures, usually calling for simple, static calculations. Where more vulnerable facilities are involved, the failure of which, or of parts of which, could subject the environment to harmful substances, more sophisticated methods are used to compute or verify their design, often accompanied by safety margins intended to compensate for uncertainties encountered at various stages of the analysis that begins with input seismic data and culminates with an effective anti-seismic design. The forthcoming discussion will deal with what is known of the characteristics of strong ground motion, highly variable according to context, without entering into the problems raised by seismotectonic studies, which actually constitute the first aspect that must be addressed when performing such an analysis. Our conclusion will be devoted to cogent R and D work in this area

  9. 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)

  10. 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 ...

  11. Infusing Reliability Techniques into Software Safety Analysis

    Science.gov (United States)

    Shi, Ying

    2015-01-01

    Software safety analysis for a large software intensive system is always a challenge. Software safety practitioners need to ensure that software related hazards are completely identified, controlled, and tracked. This paper discusses in detail how to incorporate the traditional reliability techniques into the entire software safety analysis process. In addition, this paper addresses how information can be effectively shared between the various practitioners involved in the software safety analyses. The author has successfully applied the approach to several aerospace applications. Examples are provided to illustrate the key steps of the proposed approach.

  12. 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

  13. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    systems important to safety in nuclear power plants, for all phases of the system life cycle. The guidance is applicable to systems important to safety. Since at present the reliability of a computer based system cannot be predicted on the sole basis of, or built in by, the design process, it is difficult to define and to agree systematically on any possible relaxation in the guidance to apply to software for safety related systems. Whenever possible, recommendations which apply only to safety systems and not to safety related systems are explicitly identified. The guidance relates primarily to the software used in computer based systems important to safety. Guidance on the other aspects of computer based systems, such as those concerned with the design of the computer based system itself and its hardware, is limited to the issues raised by the development, verification and validation of software.The main focus of this Safety Guide is on the preparation of documentation that is used for an adequate demonstration of the safety and reliability of computer based systems important to safety.This Safety Guide applies to all types of software: pre-existing software or firmware (such as an operating system), software to be specifically developed for the project, or software to be developed from an existing pre developed equipment family of hardware or software modules. This Safety Guide is intended for use by those involved in the production, assessment and licensing of computer based systems, including plant system designers, software designers and programmers, verifiers, validators, certifiers and regulators, as well as plant operators. The various interfaces between those involved are considered. (author)

  14. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    systems important to safety in nuclear power plants, for all phases of the system life cycle. The guidance is applicable to systems important to safety. Since at present the reliability of a computer based system cannot be predicted on the sole basis of, or built in by, the design process, it is difficult to define and to agree systematically on any possible relaxation in the guidance to apply to software for safety related systems. Whenever possible, recommendations which apply only to safety systems and not to safety related systems are explicitly identified. The guidance relates primarily to the software used in computer based systems important to safety. Guidance on the other aspects of computer based systems, such as those concerned with the design of the computer based system itself and its hardware, is limited to the issues raised by the development, verification and validation of software.The main focus of this Safety Guide is on the preparation of documentation that is used for an adequate demonstration of the safety and reliability of computer based systems important to safety. This Safety Guide applies to all types of software: pre-existing software or firmware (such as an operating system), software to be specifically developed for the project, or software to be developed from an existing pre developed equipment family of hardware or software modules. This Safety Guide is intended for use by those involved in the production, assessment and licensing of computer based systems, including plant system designers, software designers and programmers, verifiers, validators, certifiers and regulators, as well as plant operators. The various interfaces between those involved are considered

  15. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    systems important to safety in nuclear power plants, for all phases of the system life cycle. The guidance is applicable to systems important to safety. Since at present the reliability of a computer based system cannot be predicted on the sole basis of, or built in by, the design process, it is difficult to define and to agree systematically on any possible relaxation in the guidance to apply to software for safety related systems. Whenever possible, recommendations which apply only to safety systems and not to safety related systems are explicitly identified. The guidance relates primarily to the software used in computer based systems important to safety. Guidance on the other aspects of computer based systems, such as those concerned with the design of the computer based system itself and its hardware, is limited to the issues raised by the development, verification and validation of software.The main focus of this Safety Guide is on the preparation of documentation that is used for an adequate demonstration of the safety and reliability of computer based systems important to safety.This Safety Guide applies to all types of software: pre-existing software or firmware (such as an operating system), software to be specifically developed for the project, or software to be developed from an existing pre developed equipment family of hardware or software modules. This Safety Guide is intended for use by those involved in the production, assessment and licensing of computer based systems, including plant system designers, software designers and programmers, verifiers, validators, certifiers and regulators, as well as plant operators. The various interfaces between those involved are considered

  16. Nuclear safety guide: TID--7016, Revision 2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-01-01

    The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. Revision 2 of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams experienced in the field. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the former Guides by employing appropriate safety factors

  17. IAEA activity related to safety of nuclear desalination

    International Nuclear Information System (INIS)

    Gasparini, M.

    2000-01-01

    The nuclear plants for desalination to be built in the future will have to meet the standards of safety required for the best nuclear power plants currently in operation or being designed. The current safety approach, based on the achievement of the fundamental safety functions and defence in depth strategy, has been shown to be a sound foundation for the safety and protection of public health, and gives the plant the capability of dealing with a large variety of sequences, even beyond the design basis. The Department of Nuclear Safety of the IAEA is involved in many activities, the most important of which are to establish safety standards, and to provide various safety services and technical knowledge in many Technical Co-operation assistance projects. The department is also involved in other safety areas, notably in the field of future reactors. The IAEA is carrying out a project on the safety of new generation reactors, including those used for desalination, with the objective of fostering an exchange of information on safety approaches, promoting harmonization among Member States and contributing towards the development and revision of safety standards and guidelines for nuclear power plant design. The safety, regulatory and environmental concerns in nuclear powered desalination are those related directly to nuclear power plants, with due consideration given to the coupling process. The protection of product water against radioactive contamination must be ensured. An effective infrastructure, including appropriate training, a legal framework and regulatory regime, is a prerequisite to considering use of nuclear power for desalination plants, also in those countries with limited industrial infrastructures and little experience in nuclear technology or safety. (author)

  18. Nuclear Criticality Safety Assessment Using the SCALE Computer Code Package. A demonstration based on an independent review of a real application

    International Nuclear Information System (INIS)

    Mennerdahl, Dennis

    1998-06-01

    The purpose of this project was to instruct a young scientist from the Lithuanian Energy Institute (LEI) on how to carry out an independent review of a safety report. In particular, emphasis, was to be put on how to use the personal computer version of the calculation system SCALE 4.3 in this process. Nuclear criticality safety together with radiation shielding from gamma and neutron sources were areas of interest. This report concentrates on nuclear criticality safety aspects while a separate report covers radiation shielding. The application was a proposed storage cask for irradiated fuel assemblies from the Ignalina RBMK reactors in Lithuania. The safety report contained various documents involving many design and safety considerations. A few other documents describing the Ignalina reactors and their operation were available. The time for the project was limited to approximately one month, starting 'clean' with a SCALE 4.3 CD-ROM, a thick safety report and a fast personal computer. The results should be of general interest to Swedish authorities, in particular related to shielding where experience in using advanced computer codes like those available in SCALE is limited. It has been known for many years that criticality safety is very complicated, and that independent reviews are absolutely necessary to reduce the risk from quite common errors in the safety assessments. Several important results were obtained during the project. Concerning use of SCALE 4.3, it was confirmed that a young scientist, without extensive previous experience in the code system, can learn to use essentially all options. During the project, it was obvious that familiarity with personal computers, operating systems (including network system) and office software (word processing, spreadsheet and Internet browser software) saved a lot of time. Some of the Monte Carlo calculations took several hours. Experience is valuable in quickly picking out input or source document errors. Understanding

  19. Mining Behavior Based Safety Data to Predict Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey C. Joe

    2010-06-01

    The Idaho National Laboratory (INL) operates a behavior based safety program called Safety Observations Achieve Results (SOAR). This peer-to-peer observation program encourages employees to perform in-field observations of each other's work practices and habits (i.e., behaviors). The underlying premise of conducting these observations is that more serious accidents are prevented from occurring because lower level “at risk” behaviors are identified and corrected before they can propagate into culturally accepted “unsafe” behaviors that result in injuries or fatalities. Although the approach increases employee involvement in safety, the premise of the program has not been subject to sufficient empirical evaluation. The INL now has a significant amount of SOAR data on these lower level “at risk” behaviors. This paper describes the use of data mining techniques to analyze these data to determine whether they can predict if and when a more serious accident will occur.

  20. 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)

  1. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

    2001-01-01

    A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

  2. A collision model for safety evaluation of autonomous intelligent cruise control.

    Science.gov (United States)

    Touran, A; Brackstone, M A; McDonald, M

    1999-09-01

    This paper describes a general framework for safety evaluation of autonomous intelligent cruise control in rear-end collisions. Using data and specifications from prototype devices, two collision models are developed. One model considers a train of four cars, one of which is equipped with autonomous intelligent cruise control. This model considers the car in front and two cars following the equipped car. In the second model, none of the cars is equipped with the device. Each model can predict the possibility of rear-end collision between cars under various conditions by calculating the remaining distance between cars after the front car brakes. Comparing the two collision models allows one to evaluate the effectiveness of autonomous intelligent cruise control in preventing collisions. The models are then subjected to Monte Carlo simulation to calculate the probability of collision. Based on crash probabilities, an expected value is calculated for the number of cars involved in any collision. It is found that given the model assumptions, while equipping a car with autonomous intelligent cruise control can significantly reduce the probability of the collision with the car ahead, it may adversely affect the situation for the following cars.

  3. Uncertainty analysis for Ulysses safety evaluation report

    International Nuclear Information System (INIS)

    Frank, M.V.

    1991-01-01

    As part of the effort to review the Ulysses Final Safety Analysis Report and to understand the risk of plutonium release from the Ulysses spacecraft General Purpose Heat Source---Radioisotope Thermal Generator (GPHS-RTG), the Interagency Nuclear Safety Review Panel (INSRP) and the author performed an integrated, quantitative analysis of the uncertainties of the calculated risk of plutonium release from Ulysses. Using state-of-art probabilistic risk assessment technology, the uncertainty analysis accounted for both variability and uncertainty of the key parameters of the risk analysis. The results show that INSRP had high confidence that risk of fatal cancers from potential plutonium release associated with calculated launch and deployment accident scenarios is low

  4. Nuclear safety research in France

    International Nuclear Information System (INIS)

    Tanguy, P.

    1976-01-01

    As a consequence of the decision of choosing light water reactors (PWR) for the French nuclear plants of the next ten years, a large safety program has been launched referring to three physical barriers against fission product release: the fuel element cladding, main primary system boundary and the containment. The parallel development of French-designed fast breeder reactors involved safety studies on: sodium boiling, accidental fuel behavior, molten fuel-sodium interaction, core accident and protection, and external containment. The rapid development of nuclear energy resulted in a corresponding development of safety studies relating to nuclear fuel facilities. French regulations also required a special program to be developed for the realistic evaluation of the consequences of external agressions, the French cooperation to multinational safety research being also intensive

  5. Safety Culture Assessment at Regulatory Body - PNRA Experience of Implementing IAEA Methodology for Safety Culture Self Assessment

    International Nuclear Information System (INIS)

    Bhatti, S.A.N.; Arshad, N.

    2016-01-01

    The prevalence of a good safety culture is equally important for all kind of organizations involved in nuclear business including operating organizations, designers, regulator, etc., and this should be reflected through all the processes and activities of these organizations. The need for inculcating safety culture into regulatory processes and practices is gradually increasing since the major accident at Fukushima. Accordingly, several international fora in last few years repeatedly highlighted the importance of prevalence of safety culture in regulatory bodies as well. The utilisation of concept of safety culture always remained applicable in regulatory activities of PNRA in the form of core values. After the Fukushima accident, PNRA considered it important to check the extent of utilisation of safety culture concept in organizational activities and decided to conduct its “Safety Culture Self-Assessment (SCSA)” for presenting itself as a role model in-order to endorse the fact that safety culture at regulatory authority plays an important role to influence safety culture at licenced facilities.

  6. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  7. Regulatory practices of radiation safety of SNF transportation in Russia

    International Nuclear Information System (INIS)

    Kuryndina, Lidia; Kuryndin, Anton; Stroganov, Anatoly

    2008-01-01

    This paper overviews current regulatory practices for the assurance of nuclear and radiation safety during railway transportation of SNF on the territory of Russian Federation from NPPs to longterm-storage of reprocessing sites. The legal and regulatory requirements (mostly compliant with IAEA ST-1), licensing procedure for NM transportation are discussed. The current procedure does not require a regulatory approval for each particular shipment if the SNF fully comply with the Rosatom's branch standard and is transported in approved casks. It has been demonstrated that SNF packages compliant with the branch standard, which is knowingly provide sufficient safety margin, will conform to the federal level regulations. The regulatory approval is required if a particular shipment does not comply with the branch standard. In this case, the shipment can be approved only after regulatory review of Applicant's documents to demonstrate that the shipment still conformant to the higher level (federal) regulations. The regulatory review frequently needs a full calculation test of the radiation safety assurance. This test can take a lot of time. That's why the special calculation tools were created in SEC NRS. These tools aimed for precision calculation of the radiation safety parameters by SNF transportation use preliminary calculated Green's functions. Such approach allows quickly simulate any source distribution and optimize spent fuel assemblies placement in cask due to the transport equation property of linearity relatively the source. The short description of calculation tools are presented. Also, the paper discusses foreseen implications related to transportation of mixed-oxide SNF. (author)

  8. The Prospect of Motorcycle Safety Education in Secondary Schools.

    Science.gov (United States)

    King, Alfred S.

    Motorcycle safety education will become a necessity in the near future due to the growing demands of secondary students for education in this area. The Motorcycle Safety Foundation is sponsored by major motorcycle industries and is involved with developing programs and materials to promote motorcycle safety education. The high rate of motorcycle…

  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. Experience of RIA safety analyses performance for NPP Temelin core arranged with TVSA-T fuel assemblies

    International Nuclear Information System (INIS)

    Kryukov, S.A.; Lizorkin, M.P.

    2010-01-01

    The contents of the presentation are as follows: 1. Definition of categories for initiating events; 2. Acceptance criteria for safety assessment; 3. Main aspects of safety assessment methodology; 4. Main stages of calculation analysis; 5. Interface with other parts of the core design; 6. Codes used for calculation; 6.1 Main performances of code package TIGR-1; 6.2 Main performances of code BIPR-7A; 7. TIGR-1 accounting of design margins in calculation of fuel rod powers; 8. Peculiar features of Instrumentation and Control System for Temelin NPP; 9. Calculations; 10. Checklist of margin data important for reload safety assessment. (P.A.)

  11. Safety issues at the defense production reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors endash the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. The report is necessarily based on a limited review of the defense production reactors. It does not address whether any of the reactors are ''safe,'' because such an analysis would involve a determination of acceptable risk endash a matter of obvious importance, but one that was beyond the purview of the committee. It also does not address whether the safety of the production reactors is comparable to that of commercial nuclear power stations, because even this narrower question extended beyond the charge to the committee and would have involved detailed analyses that the committee could not undertake

  12. Safety reviews of the Brazilian multipurpose reactor

    International Nuclear Information System (INIS)

    Soares, Humberto Vitor

    2014-01-01

    This work presents a model developed for thermal hydraulic (TH) simulation of the Multipurpose Brazilian Reactor (RMB), whose Brazilian proposal for design, construction and operation was established in 2007. This reactor has as main proposed the production of radioisotopes for use in exams of nuclear medicine, material tests and utilization of neutrons beams. Besides of the TH modeling and safety analysis of the reactor, the application of a methodology to perform coupled calculation thermal-hydraulic/neutron kinetic (TH/NK) is also presented. Initially, the RMB was modeled in the safety analysis RELAP5 code. This code performs the thermal hydraulic calculation using point kinetics. Subsequently, the model was adapted and verified to the RELAP5-3D© code. This code performs the process of internal coupling through the option of nodal neutron kinetics calculation using the NESTLE code which solves the neutron diffusion equation. To generate the neutronic group constants, which are macroscopic cross sections that serve as input data for the neutronic codes, it was used the WIMSD-5B cell calculation code. The neutron analysis code PARCS was also used to model the 3D RMB core in order to compare the results of radial and axial average power distribution with the results generated by RELAP5-3D© code and with the available results of the CITATION neutron kinetic code. The safety analyses demonstrated safe behavior of the reactor through situations of possible transients. The 3D coupled calculations to the steady state operation also showed expected behavior, as well as the RMB neutronic analyzes performed with the codes NESTLE and PARCS.(author)

  13. 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.

  14. 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.

  15. 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.

  16. Reliability Analysis for Safety Grade PLC(POSAFE-Q)

    International Nuclear Information System (INIS)

    Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

    2012-01-01

    Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

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

    International Nuclear Information System (INIS)

    2003-01-01

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

  18. Nuclear safety guide. TID-7016, Revision 2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-01-01

    The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. The present revision of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams, experienced in the field. The reader will find a significant change in the character of information presented in this version. Nuclear Criticality Safety has matured in the past twelve years. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the formerGuides by employing appropriate safety factors. In fact, it becomes incumbent on the Criticality Safety Specialist to necessarily impose safety factors consistent with the possible normal and abnormal credible contingencies of an operation as revealed by his evaluation. In its present form the Guide easily becomes a suitable module in any compendium or handbook tailored for internal use by organizations. It is hoped the Guide will continue to serve immediate needs and will encourage continuing and more comprehensive efforts toward organizing nuclear criticality safety information

  19. Nuclear safety guide. TID-7016, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, J T [ed.

    1978-05-01

    The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. The present revision of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams, experienced in the field. The reader will find a significant change in the character of information presented in this version. Nuclear Criticality Safety has matured in the past twelve years. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the formerGuides by employing appropriate safety factors. In fact, it becomes incumbent on the Criticality Safety Specialist to necessarily impose safety factors consistent with the possible normal and abnormal credible contingencies of an operation as revealed by his evaluation. In its present form the Guide easily becomes a suitable module in any compendium or handbook tailored for internal use by organizations. It is hoped the Guide will continue to serve immediate needs and will encourage continuing and more comprehensive efforts toward organizing nuclear criticality safety information.

  20. Bias in calculated keff from subcritical measurements by the 252Cf-source-driven noise analysis method

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Valentine, T.E.

    1995-01-01

    The development of MCNP-DSP, which allows direct calculation of the measured time and frequency analysis parameters from subcritical measurements using the 252 Cf-source-driven noise analysis method, permits the validation of calculational methods for criticality safety with in-plant subcritical measurements. In addition, a method of obtaining the bias in the calculations, which is essential to the criticality safety specialist, is illustrated using the results of measurements with 17.771-cm-diam, enriched (93.15), unreflected, and unmoderated uranium metal cylinders. For these uranium metal cylinders the bias obtained using MCNP-DSP and ENDF/B-V cross-section data increased with subcriticality. For a critical experiment [height (h) = 12.629 cm], it was -0.0061 ± 0.0003. For a 10.16-cm-high cylinder (k ∼ 0.93), it was 0.0060 ± 0.0016, and for a subcritical cylinder (h = 8.13 cm, k ∼ 0.85), the bias was -0.0137 ± 0.0037, more than a factor of 2 larger in magnitude. This method allows the nuclear criticality safety specialist to establish the bias in calculational methods for criticality safety from in-plant subcritical measurements by the 252 Cf-source-driven noise analysis method

  1. Architecture Level Safety Analyses for Safety-Critical Systems

    Directory of Open Access Journals (Sweden)

    K. S. Kushal

    2017-01-01

    Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

  2. 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

  3. A calculational system to aid economical use of MTRs

    International Nuclear Information System (INIS)

    Reitsma, F.; Joubert, W.R.

    1999-01-01

    The availability of a fast and accurate core neutronic calculational system is a valuable asset in the operation and utilization of research reactors. Its primary value lies in optimum reload design, fuel management, safety and utilization studies. In this paper the OSCAR-3 calculational system of the Atomic Energy Corporation of South Africa (AEC) is discussed in detail. The different components and important features are explained with a short summary of some comparisons with experiments. (author)

  4. Perceived community environment and physical activity involvement in a northern-rural Aboriginal community

    Directory of Open Access Journals (Sweden)

    Lévesque Lucie

    2007-12-01

    Full Text Available Abstract Background Type 2 diabetes disproportionately affects Aboriginal peoples in Canada. Ample evidence shows that regular physical activity (PA plays an important role in the prevention and treatment of type 2 diabetes. Evidence is beginning to emerge linking PA to the physical environment but little is known about the relationship between remote rural environments and PA involvement in Aboriginal peoples. This study's purpose was to investigate the relationship between perceptions of the environment and PA and walking patterns in Aboriginal adults in order to inform the planning and implementation of community-relevant PA interventions. Methods Two hundred and sixty three residents (133 women, mean age = 35.6 years, SD = 12.3 and 130 men, mean age = 37.2 years, SD = 13.1 from Moose Factory, Ontario were asked about environmental factors related to walking and PA involvement. Survey items were drawn from standardized, validated questionnaires. Descriptive statistics (means, standard deviations, percentages were calculated. A series of hierarchical multiple regressions were performed to determine associations between walking and overall PA with perceived environmental variables. Results Hierarchical multiple regression to predict walking revealed significant associations between walking and perceived safety and aesthetics. Owning home exercise equipment predicted strenuous PA. Different aspects of the physical environment appear to influence different types of physical activities. The significant amount of variance in behaviour accounted for by perceived environmental variables (5.3% walking included safety, aesthetics, convenience, owning home exercise equipment and comfortable shoes for walking. Conclusion Results suggest that a supportive physical environment is important for PA involvement and that walking and activities of different intensity appear to be mediated by different perceived environmental variables. Implications for PA

  5. 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)

  6. 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)

  7. Nuclear Criticality Safety Data Book

    Energy Technology Data Exchange (ETDEWEB)

    Hollenbach, D. F. [Y-12 National Security Complex, Oak Ridge, TN (United States)

    2016-11-14

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  8. Nuclear Criticality Safety Data Book

    International Nuclear Information System (INIS)

    Hollenbach, D. F.

    2016-01-01

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  9. 75 FR 54804 - Safety and Health Management Programs for Mines

    Science.gov (United States)

    2010-09-09

    .... Worker Involvement. 3. Hazard Identification, including workplace inspections for violations of mandatory... requirements; it reflects the embodiment of a culture of safety--from the CEO to the worker to the contractor. This culture of safety derives from a commitment to a systematic, effective, comprehensive safety and...

  10. Improving construction site safety through leader-based verbal safety communication.

    Science.gov (United States)

    Kines, Pete; Andersen, Lars P S; Spangenberg, Soren; Mikkelsen, Kim L; Dyreborg, Johnny; Zohar, Dov

    2010-10-01

    . Coaching construction site foremen to include safety in their daily verbal exchanges with workers has a significantly positive and lasting effect on the level of safety, which is a proximal estimate for work-related accidents. It is recommended that future studies include coaching and feedback at all organizational levels and for all involved parties in the construction process. Building client regulations could assign the task of coaching to the client appointed safety coordinators or a manager/supervisor, and studies should measure longitudinal effects of coaching by following foremen and their work gangs from site to site. Copyright © 2010 National Safety Council and Elsevier Ltd. Published by Elsevier Ltd. All rights reserved.

  11. 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.

  12. Road safety issues for bus transport management.

    Science.gov (United States)

    Cafiso, Salvatore; Di Graziano, Alessandro; Pappalardo, Giuseppina

    2013-11-01

    Because of the low percentage of crashes involving buses and the assumption that public transport improves road safety by reducing vehicular traffic, public interest in bus safety is not as great as that in the safety of other types of vehicles. It is possible that less attention is paid to the significance of crashes involving buses because the safety level of bus systems is considered to be adequate. The purpose of this study was to evaluate the knowledge and perceptions of bus managers with respect to safety issues and the potential effectiveness of various technologies in achieving higher safety standards. Bus managers were asked to give their opinions on safety issues related to drivers (training, skills, performance evaluation and behaviour), vehicles (maintenance and advanced devices) and roads (road and traffic safety issues) in response to a research survey. Kendall's algorithm was used to evaluate the level of concordance. The results showed that the majority of the proposed items were considered to have great potential for improving bus safety. The data indicated that in the experience of the participants, passenger unloading and pedestrians crossing near bus stops are the most dangerous actions with respect to vulnerable users. The final results of the investigation showed that start inhibition, automatic door opening, and the materials and internal architecture of buses were considered the items most strongly related to bus passenger safety. Brake assistance and vehicle monitoring systems were also considered to be very effective. With the exception of driver assistance systems for passenger and pedestrian safety, the perceptions of the importance of other driver assistance systems for vehicle monitoring and bus safety were not unanimous among the bus company managers who participated in this survey. The study results showed that the introduction of new technologies is perceived as an important factor in improving bus safety, but a better understanding

  13. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition); Seguridad de las centrales nucleares: Diseno. Requisitos de seguridad especificos

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  14. Nuclear criticality safety handbook. Version 2

    International Nuclear Information System (INIS)

    1999-03-01

    The Nuclear Criticality Safety Handbook, Version 2 essentially includes the description of the Supplement Report to the Nuclear Criticality Safety Handbook, released in 1995, into the first version of Nuclear Criticality Safety Handbook, published in 1988. The following two points are new: (1) exemplifying safety margins related to modelled dissolution and extraction processes, (2) describing evaluation methods and alarm system for criticality accidents. Revision is made based on previous studies for the chapter that treats modelling the fuel system: e.g., the fuel grain size that the system can be regarded as homogeneous, non-uniformity effect of fuel solution, and burnup credit. This revision solves the inconsistencies found in the first version between the evaluation of errors found in JACS code system and criticality condition data that were calculated based on the evaluation. (author)

  15. FAKIR: a user-friendly standard for decay heat and activity calculation of LWR fuel

    International Nuclear Information System (INIS)

    Pretesacque, P.; Nimal, J.C.; Huynh, T.D.; Zachar, M.

    1993-01-01

    The shipping casks owned by the transporters and the unloading and storage facilities are subjected by their design safety report to decay heat and activity limits. It is the responsibility of the consignor or the consignee to check the compliance of the fuel assemblies to the shipped or stored with regard to these limiting safety parameters. Considering the diversity of the parties involved in the transport and storage cycle, a standardization has become necessary. This has been achieved by the FAKIR code. The FAKIR development started in 1984 in collaboration between COGEMA, CEA-SERMA and NTL. Its main specifications were to be a user-friendly code, to use the contractual data given in the COGEMA transport and reprocessing sheet 1 as input, and to over-estimate decay heat and activity. Originally based on computerizable standards such as ANSI or USNRC, the FAKIR equations and data libraries are now based on the fully qualified PEPIN/APOLLO calculation codes. FAKIR is applicable to all patterns of irradiation histories, with burn up from 1000 MWd/TeU to 70.000 MWd/TeU and cooling times from 1 second to 100 years. (J.P.N.)

  16. DCHAIN-SP 2001: High energy particle induced radioactivity calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)

    2001-03-01

    For the purpose of contribution to safety design calculations for induced radioactivities in the JAERI/KEK high-intensity proton accelerator project facilities, the DCHAIN-SP which calculates the high energy particle induced radioactivity has been updated to DCHAIN-SP 2001. The following three items were improved: (1) Fission yield data are included to apply the code to experimental facility design for nuclear transmutation of long-lived radioactive waste where fissionable materials are treated. (2) Activation cross section data below 20 MeV are revised. In particular, attentions are paid to cross section data of materials which have close relation to the facilities, i.e., mercury, lead and bismuth, and to tritium production cross sections which are important in terms of safety of the facilities. (3) User-interface for input/output data is sophisticated to perform calculations more efficiently than that in the previous version. Information needed for use of the code is attached in Appendices; the DCHAIN-SP 2001 manual, the procedures of installation and execution of DCHAIN-SP, and sample problems. (author)

  17. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  18. Status of safety issues at licensed power plants: TMI action plan requirements, unresolved safety issues, generic safety issues

    International Nuclear Information System (INIS)

    1991-12-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program was established whereby an annual NUREG report would be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was compiled and reported in three NUREG volumes. Volume 1, published in March 1991, addressed the status of of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). This annual NUREG report combines these volumes into a single report and provides updated information as of September 30, 1991. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. This report is to provide a comprehensive description of the implementation and verification status of TMI Action Plan Requirements, safety issues designated as USIs, and GSIs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

  19. 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)

  20. The effect of terrain slope on firefighter safety zone effectiveness

    Science.gov (United States)

    Bret Butler; J. Forthofer; K. Shannon; D. Jimenez; D. Frankman

    2010-01-01

    The current safety zone guidelines used in the US were developed based on the assumption that the fire and safety zone were located on flat terrain. The minimum safe distance for a firefighter to be from a flame was calculated as that corresponding to a radiant incident energy flux level of 7.0kW-m-2. Current firefighter safety guidelines are based on the assumption...

  1. 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 ...

  2. 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

  3. Nulcear Safety: Technical progress review, October--December 1988

    Energy Technology Data Exchange (ETDEWEB)

    Silver, E G [ed.

    1988-01-01

    Nuclear Safety 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.

  4. Improving method for calculating integral index of personnel security of company

    Directory of Open Access Journals (Sweden)

    Chjan Khao Yui

    2016-06-01

    Full Text Available The paper improves the method of calculating the integral index of personnel security of a company. The author has identified four components of personnel security (social and motivational safety, occupational safety, not confliction security, life safety which are characterized by certain indicators. Integral index of personnel security is designed for the enterprises of machine-building sector in Kharkov region, taking into account theweight coefficients j-th component of bj, and weighting factors that determine the degree of contribution of the ith parameter in the integral index aіj as defined by experts.

  5. 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.

  6. Regulatory role and approach of BARC Safety Council in safety and occupational health in BARC facilities

    International Nuclear Information System (INIS)

    Rajdeep; Jayarajan, K.; Taly, Y.K.

    2016-01-01

    Bhabha Atomic Research Centre is involved in multidisciplinary research and developmental activities, related to peaceful use of nuclear energy and its societal benefits. In order to achieve high level of performance of these facilities, the best efforts are made to maintain good health of the plant personnel and good working conditions. BARC Safety Council (BSC), which is the regulatory body for BARC facilities, regulates radiation safety, industrial safety and surveillance of occupational health, by implementing various rules and guidelines in BARC facilities. BARC Safety framework consists of various committees in a 3-tier system. The first tier is BSC, which is the apex body authorized for issuing directives, permissions, consents and authorizations. It is having responsibility of ensuring protection and safety of public, environment, personnel and facilities of BARC through enforcement of radiation protection and industrial safety programmes. Besides the 18 committees in 2"n"d tier, there are 6 other expert committees which assist in functioning of BSC. (author)

  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. Criticality Safety Evaluation of Hanford Tank Farms Facility

    Energy Technology Data Exchange (ETDEWEB)

    WEISS, E.V.

    2000-12-15

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

  9. Criticality Safety Evaluation of Hanford Tank Farms Facility

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste

  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. 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

  12. Safety culture: personal considerations of an owner/operator

    International Nuclear Information System (INIS)

    Fuchs, H.

    1994-01-01

    Safety culture with nuclear energy is above all a people's business. This means that all you can do is attempting to create the type of ideal environment that helps all plant people to perform their safety-related tasks in an optimum way. This is a continuous challenge for all who are involved. In the last years the political environment has exhibited the most noteworthy shortcomings regarding safety culture. (author) figs

  13. 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

  14. 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

  15. 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

  16. Calculation code evaluating the confinement of a nuclear facility in case of fires

    International Nuclear Information System (INIS)

    Laborde, J.C.; Prevost, C.; Vendel, J.

    1995-01-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation

  17. Calculation code evaluating the confinement of a nuclear facility in case of fires

    Energy Technology Data Exchange (ETDEWEB)

    Laborde, J.C.; Prevost, C.; Vendel, J. [and others

    1995-02-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.

  18. Patient participation in patient safety and nursing input - a systematic review.

    Science.gov (United States)

    Vaismoradi, Mojtaba; Jordan, Sue; Kangasniemi, Mari

    2015-03-01

    This systematic review aims to synthesise the existing research on how patients participate in patient safety initiatives. Ambiguities remain about how patients participate in routine measures designed to promote patient safety. Systematic review using integrative methods. Electronic databases were searched using keywords describing patient involvement, nursing input and patient safety initiatives to retrieve empirical research published between 2007 and 2013. Findings were synthesized using the theoretical domains of Vincent's framework for analysing risk and safety in clinical practice: "patient", "healthcare provider", "task", "work environment", "organisation & management". We identified 17 empirical research papers: four qualitative, one mixed-method and 12 quantitative designs. All 17 papers indicated that patients can participate in safety initiatives. Improving patient participation in patient safety necessitates considering the patient as a person, the nurse as healthcare provider, the task of participation and the clinical environment. Patients' knowledge, health conditions, beliefs and experiences influence their decisions to engage in patient safety initiatives. An important component of the management of long-term conditions is to ensure that patients have sufficient knowledge to participate. Healthcare providers may need further professional development in patient education and patient care management to promote patient involvement in patient safety, and ensure that patients understand that they are 'allowed' to inform nurses of adverse events or errors. A healthcare system characterised by patient-centredness and mutual acknowledgement will support patient participation in safety practices. Further research is required to improve international knowledge of patient participation in patient safety in different disciplines, contexts and cultures. Patients have a significant role to play in enhancing their own safety while receiving hospital care. This

  19. 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

  20. Patient and family involvement in contemporary health care.

    Science.gov (United States)

    Angood, Peter; Dingman, Jennifer; Foley, Mary E; Ford, Dan; Martins, Becky; O'Regan, Patti; Salamendra, Arlene; Sheridan, Sue; Denham, Charles R

    2010-03-01

    The objective of this article was to provide a guide to health care providers on patient and family involvement in health care. This article evaluated the latest published studies for patient and family involvement and reexamined the objectives, the requirements for achieving these objectives, and the evidence of how to involve patients and families. Critical components for patient safety include changing the organizational culture; including patients and families on teams; listening to patients and families; incorporating their input into leadership structures and systems; providing full detail about treatment, procedures, and medication adverse effects; involving them on patient safety and performance improvement committees; and disclosing medical errors. The conclusion of this article is that, for the future, patient and family involvement starts with educating patients and families and ends with listening to them and taking them seriously. If patient and family input is emphatically built into systems of performance improvement, and if patients and families are taken seriously and are respected for their valuable perspectives about how care can be improved, then organizations can improve at improving. Resources in health care are in short supply, yet the resources of patient and family help and time are almost limitless, are ready to be tapped, and can have a huge impact on improving the reliability and overall success for any health care organization.

  1. A university contribution to reactor safety

    International Nuclear Information System (INIS)

    Hall, W.B.

    1980-01-01

    The total UK university effort available for research specifically directed towards reactor safety is certainly small in comparison with that in industry. To be worth while, the work should complement that in the industry, and ways in which this can, and in some cases does, happen, will be discussed. There is, however, another reason for university involvement: the need for an informed body of opinion on matters of reactor safety outside the industry. Without this it is difficult for the public and its representatives to assure themselves that the depth and scope of safety analysis is commensurate with the seriousness of the problem, and that the best available data and techniques are being used. An independent inspectorate is an essential element in this philosophy, but in addition there is much to be said for exposing the arguments to scrutiny by the widest possible range of informed critics. Such people will be much more effective if they are themselves involved in real problems in the field. In a university, this involvement is probably best achieved through research; as mentioned above, the type of research should preferably complement that being carried out in the industry. The current situation, and the future, are discussed. (author)

  2. Middle East food safety perspectives.

    Science.gov (United States)

    Idriss, Atef W; El-Habbab, Mohammad S

    2014-08-01

    Food safety and quality assurance are increasingly a major issue with the globalisation of agricultural trade, on the one hand, and intensification of agriculture, on the other. Consumer protection has become a priority in policy-making amongst the large economies of the Middle East and North Africa (MENA) countries following a number of food safety incidents. To enhance food safety, it is necessary to establish markets underpinned by knowledge and resources, including analysis of international rejections of food products from MENA countries, international laboratory accreditation, improved reporting systems and traceability, continued development and validation of analytical methods, and more work on correlating sensory evaluation with analytical results. MENA countries should develop a national strategy for food safety based on a holistic approach that extends from farm-to-fork and involves all the relevant stakeholders. Accordingly, food safety should be a regional programme, raising awareness among policy- and decision-makers of the importance of food safety and quality for consumer protection, food trade and economic development. © 2014 Society of Chemical Industry.

  3. Reconstruction calculation of pin power for ship reactor core

    International Nuclear Information System (INIS)

    Li Haofeng; Shang Xueli; Chen Wenzhen; Wang Qiao

    2010-01-01

    Aiming at the limitation of the software that pin power distribution for ship reactor core was unavailable, the calculation model and method of the axial and radial pin power distribution were proposed. Reconstruction calculations of pin power along axis and radius was carried out by bicubic and bilinear interpolation and cubic spline interpolation, respectively. The results were compared with those obtained by professional reactor physical soft with fine mesh difference. It is shown that our reconstruction calculation of pin power is simple and reliable as well as accurate, which provides an important theoretic base for the safety analysis and operating administration of the ship nuclear reactor. (authors)

  4. Compilation report of VHTRC temperature coefficient benchmark calculations

    Energy Technology Data Exchange (ETDEWEB)

    Yasuda, Hideshi; Yamane, Tsuyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-11-01

    A calculational benchmark problem has been proposed by JAERI to an IAEA Coordinated Research Program, `Verification of Safety Related Neutronic Calculation for Low-enriched Gas-cooled Reactors` to investigate the accuracy of calculation results obtained by using codes of the participating countries. This benchmark is made on the basis of assembly heating experiments at a pin-in block type critical assembly, VHTRC. Requested calculation items are the cell parameters, effective multiplication factor, temperature coefficient of reactivity, reaction rates, fission rate distribution, etc. Seven institutions from five countries have joined the benchmark works. Calculation results are summarized in this report with some remarks by the authors. Each institute analyzed the problem by applying the calculation code system which was prepared for the HTGR development of individual country. The values of the most important parameter, k{sub eff}, by all institutes showed good agreement with each other and with the experimental ones within 1%. The temperature coefficient agreed within 13%. The values of several cell parameters calculated by several institutes did not agree with the other`s ones. It will be necessary to check the calculation conditions again for getting better agreement. (J.P.N.).

  5. Fire Safety Trianing in Health Care Institutions.

    Science.gov (United States)

    American Hospital Association, Chicago, IL.

    The manual details the procedures to be followed in developing and implementing a fire safety plan. The three main steps are first, to organize; second, to set up a procedure and put it in writing; and third, to train and drill employees and staff. Step 1 involves organizing a safety committee, appointing a fire marshall, and seeking help from…

  6. Nuclear materials facility safety initiative

    International Nuclear Information System (INIS)

    Peddicord, K.L.; Nelson, P.; Roundhill, M.; Jardine, L.J.; Lazarev, L.; Moshkov, M.; Khromov, V.V.; Kruchkov, E.; Bolyatko, V.; Kazanskij, Yu.; Vorobeva, I.; Lash, T.R.; Newton, D.; Harris, B.

    2000-01-01

    Safety in any facility in the nuclear fuel cycle is a fundamental goal. However, it is recognized that, for example, should an accident occur in either the U.S. or Russia, the results could seriously delay joint activities to store and disposition weapons fissile materials in both countries. To address this, plans are underway jointly to develop a nuclear materials facility safety initiative. The focus of the initiative would be to share expertise which would lead in improvements in safety and safe practices in the nuclear fuel cycle.The program has two components. The first is a lab-to-lab initiative. The second involves university-to-university collaboration.The lab-to-lab and university-to-university programs will contribute to increased safety in facilities dealing with nuclear materials and related processes. These programs will support important bilateral initiatives, develop the next generation of scientists and engineers which will deal with these challenges, and foster the development of a safety culture

  7. Regulatory measures for traffic safety

    International Nuclear Information System (INIS)

    Veerapur, R.D.; Bharambe, S.D.; Patnaik, S.K.; Tandle, A.K.; Sonawane, K.A.; Kumar, Rajesh; Venkat Subramanian, K.

    2017-01-01

    Traffic safety is an issue related to occupational safety not restricted alone to the transportation but extends beyond. BARC has many facilities spread across large area in Mumbai and outside Mumbai. BARC deploys large number of buses, mini buses, jeeps and cars for commuting its employees to reach BARC and for commuting within BARC premises. Additionally, trucks, fire tenders, trailers etc. are also deployed for transportation of materials. No moving vehicle is ever free of the possibility of involvement in an accident. Vehicular accidents and the fatalities on road are the result of inter-play of a number of factors. The vehicle population has been steadily increasing with the pace picking up significantly in recent past. Increase in vehicle population in the face of limited road space used by a large variety of traffic has heightened the need and urgency for a well-thought-out road safety. Therefore, existence of regulatory authority to regulate traffic and vehicles to ensure safety of its employees and vehicles is very essential. BARC Traffic Safety Committee (BTSC), which is the regulating body for traffic safety is responsible for ensuring overall traffic safety. (author)

  8. Safety climate in Swiss hospital units: Swiss version of the Safety Climate Survey

    Science.gov (United States)

    Gehring, Katrin; Mascherek, Anna C.; Bezzola, Paula

    2015-01-01

    Abstract Rationale, aims and objectives Safety climate measurements are a broadly used element of improvement initiatives. In order to provide a sound and easy‐to‐administer instrument for the use in Swiss hospitals, we translated the Safety Climate Survey into German and French. Methods After translating the Safety Climate Survey into French and German, a cross‐sectional survey study was conducted with health care professionals (HCPs) in operating room (OR) teams and on OR‐related wards in 10 Swiss hospitals. Validity of the instrument was examined by means of Cronbach's alpha and missing rates of the single items. Item‐descriptive statistics group differences and percentage of ‘problematic responses’ (PPR) were calculated. Results 3153 HCPs completed the survey (response rate: 63.4%). 1308 individuals were excluded from the analyses because of a profession other than doctor or nurse or invalid answers (n = 1845; nurses = 1321, doctors = 523). Internal consistency of the translated Safety Climate Survey was good (Cronbach's alpha G erman = 0.86; Cronbach's alpha F rench = 0.84). Missing rates at item level were rather low (0.23–4.3%). We found significant group differences in safety climate values regarding profession, managerial function, work area and time spent in direct patient care. At item level, 14 out of 21 items showed a PPR higher than 10%. Conclusions Results indicate that the French and German translations of the Safety Climate Survey might be a useful measurement instrument for safety climate in Swiss hospital units. Analyses at item level allow for differentiating facets of safety climate into more positive and critical safety climate aspects. PMID:25656302

  9. Fatal crashes involving large numbers of vehicles and weather.

    Science.gov (United States)

    Wang, Ying; Liang, Liming; Evans, Leonard

    2017-12-01

    Adverse weather has been recognized as a significant threat to traffic safety. However, relationships between fatal crashes involving large numbers of vehicles and weather are rarely studied according to the low occurrence of crashes involving large numbers of vehicles. By using all 1,513,792 fatal crashes in the Fatality Analysis Reporting System (FARS) data, 1975-2014, we successfully described these relationships. We found: (a) fatal crashes involving more than 35 vehicles are most likely to occur in snow or fog; (b) fatal crashes in rain are three times as likely to involve 10 or more vehicles as fatal crashes in good weather; (c) fatal crashes in snow [or fog] are 24 times [35 times] as likely to involve 10 or more vehicles as fatal crashes in good weather. If the example had used 20 vehicles, the risk ratios would be 6 for rain, 158 for snow, and 171 for fog. To reduce the risk of involvement in fatal crashes with large numbers of vehicles, drivers should slow down more than they currently do under adverse weather conditions. Driver deaths per fatal crash increase slowly with increasing numbers of involved vehicles when it is snowing or raining, but more steeply when clear or foggy. We conclude that in order to reduce risk of involvement in crashes involving large numbers of vehicles, drivers must reduce speed in fog, and in snow or rain, reduce speed by even more than they already do. Copyright © 2017 National Safety Council and Elsevier Ltd. All rights reserved.

  10. 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...

  11. Interface for safety and security of radioactive sources

    International Nuclear Information System (INIS)

    Seggane, Richard

    2016-04-01

    In facilities and activities involving use of radiation sources, safety and security measures have in common the aim of protecting human life and health and the environment. In addition, safety and security measures must be designed and implemented in an integrated manner, so that security measures do not compromise safety and safety measures do not compromise security measures. This work reviewed issues related to establishing a clear interface between safety and security of radiation sources. The Government, the Regulatory Authority and licensee/registrants and other relevant stakeholders should work together and contribute to ensure that safety and security of sources is ensured and well interfaced. A Radiotherapy facility has been used as a case study. (au)

  12. Calculation of neutron flux in the presence of a source

    International Nuclear Information System (INIS)

    Planchard, J.

    1993-09-01

    Neutron sources are introduced into the reactors to initiate the chain reaction. For safety reasons, we have to know the distribution and evolution of the flux throughout the startup phase. The flux is calculated iteratively but convergence of the process can slow down arbitrarily as we approach criticality. A calculation method is presented, with a convergence speed which does not depend on the negative reactivity when it is small. (author). 7 refs

  13. Environmental and Occupational Safety Division annual progress report for 1983

    International Nuclear Information System (INIS)

    1984-11-01

    This report presents summaries of activities conducted during 1983 in the following areas: radiation monitoring; health physics instrumentation development; environmental management; atmospheric monitoring; water monitoring; background radiation measurements; soil and grass samples; deer samples; calculation of potential radiation dose to the public; industrial safety; and operational safety

  14. 49 CFR 1106.4 - The Safety Integration Plan process.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 8 2010-10-01 2010-10-01 false The Safety Integration Plan process. 1106.4 Section 1106.4 Transportation Other Regulations Relating to Transportation (Continued) SURFACE... CONSIDERATION OF SAFETY INTEGRATION PLANS IN CASES INVOLVING RAILROAD CONSOLIDATIONS, MERGERS, AND ACQUISITIONS...

  15. A non-uniform expansion mechanical safety model of the stent.

    Science.gov (United States)

    Yang, J; Huang, N; Du, Q

    2009-01-01

    Stents have a serial unstable structure that readily leads to non-uniform expansion. Non-uniform expansion in turn creates a stent safety problem. We explain how a stent may be simplified to a serial unstable structure, and present a method to calculate the non-uniform expansion of the stent on the basis of the serial unstable structure. We propose a safety criterion based on the expansion displacement instead of the strain, and explain that the parameter Rd, the ratio of the maximum displacement of the elements to normal displacement, is meaningful to assess the safety level of the stent. We also examine how laser cutting influences non-uniform expansion. The examples illustrate how to calculate the parameter Rd to assess non-uniform expansion of the stent, and demonstrate how the laser cutting offset and strengthening coefficient of the material influence the stent expansion behaviour. The methods are valuable for assessing stent safety due to non-uniform expansion.

  16. 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.

  17. Dimensions of Safety Climate among Iranian Nurses

    Directory of Open Access Journals (Sweden)

    Z Naghavi Konjin

    2015-10-01

    Full Text Available Background: 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. Objective: To develop and validate a psychometrics scale for measuring nurses' safety climate. Methods: 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. Results: 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. Conclusion: The proposed scale can be used in identifying the needed areas to implement interventions in safety climate of nurses.

  18. Simplified probabilistic approach to determine safety factors in deterministic flaw acceptance criteria

    International Nuclear Information System (INIS)

    Barthelet, B.; Ardillon, E.

    1997-01-01

    The flaw acceptance rules in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a reliable method of evaluating the safety margins and the integrity of components led Electricite de France to launch a study to link safety factors with requested reliability. A simplified analytical probabilistic approach is developed to analyse the failure risk in Fracture Mechanics. Assuming lognormal distributions of the main random variables, it is possible considering a simple Linear Elastic Fracture Mechanics model, to determine the failure probability as a function of mean values and logarithmic standard deviations. The 'design' failure point can be analytically calculated. Partial safety factors on the main variables (stress, crack size, material toughness) are obtained in relation with reliability target values. The approach is generalized to elastic plastic Fracture Mechanics (piping) by fitting J as a power law function of stress, crack size and yield strength. The simplified approach is validated by detailed probabilistic computations with PROBAN computer program. Assuming reasonable coefficients of variations (logarithmic standard deviations), the method helps to calibrate safety factors for different components taking into account reliability target values in normal, emergency and faulted conditions. Statistical data for the mechanical properties of the main basic materials complement the study. The work involves laboratory results and manufacture data. The results of this study are discussed within a working group of the French in service inspection code RSE-M. (authors)

  19. 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)

  20. Safety parameter display system: an operator support system for enhancement of safety in Indian PHWRs

    International Nuclear Information System (INIS)

    Subramaniam, K.; Biswas, T.

    1994-01-01

    Ensuring operational safety in nuclear power plants is important as operator errors are observed to contribute significantly to the occurrence of accidents. Computerized operator support systems, which process and structure information, can help operators during both normal and transient conditions, and thereby enhance safety and aid effective response to emergency conditions. An important operator aid being developed and described in this paper, is the safety parameter display system (SPDS). The SPDS is an event-independent, symptom-based operator aid for safety monitoring. Knowledge-based systems can provide operators with an improved quality of information. An information processing model of a knowledge based operator support system (KBOSS) developed for emergency conditions using an expert system shell is also presented. The paper concludes with a discussion of the design issues involved in the use of a knowledge based systems for real time safety monitoring and fault diagnosis. (author). 8 refs., 4 figs., 1 tab

  1. Principles for decisions involving environmental and health risks

    International Nuclear Information System (INIS)

    Bengtsson, B.

    1989-01-01

    Decision making with respect to safety is becoming more and more complex. The risk involved must be taken into account together with numerous other factors such as the benefits, the uncertainties and the public perception. Can the decision maker be aided by some kind of system, general rules of thumb, or broader perspective on similar decisions? This question has been addressed in a joint Nordic project relating to nuclear power. Modern techniques for risk assessment and management have been studied and parallels drawn to such areas as offshore safety and management of genotoxic chemicals in the environment. The topics include synoptic vs. incrementalistic approaches to decision making, health hazards from radiation and genotoxic chemicals, value judgments in decision making, definitions of low risks, risk comparisons, and principles for decision making when risks are involved. (author) 47 refs

  2. 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.

  3. 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

  4. Criteria for safety-related operator actions

    International Nuclear Information System (INIS)

    Gray, L.H.; Haas, P.M.

    1983-01-01

    The Safety-Related Operator Actions (SROA) Program was designed to provide information and data for use by NRC in assessing the performance of nuclear power plant (NPP) control room operators in responding to abnormal/emergency events. The primary effort involved collection and assessment of data from simulator training exercises and from historical records of abnormal/emergency events that have occurred in operating plants (field data). These data can be used to develop criteria for acceptability of the use of manual operator action for safety-related functions. Development of criteria for safety-related operator actions are considered

  5. 77 FR 21595 - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving...

    Science.gov (United States)

    2012-04-10

    ... must be one which, if proven, would entitle the requestor/petitioner to relief. A requestor/ petitioner..., and fire modeling calculations, have been performed to demonstrate that the performance-based... may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations...

  6. Cultural Humility and Hospital Safety Culture.

    Science.gov (United States)

    Hook, Joshua N; Boan, David; Davis, Don E; Aten, Jamie D; Ruiz, John M; Maryon, Thomas

    2016-12-01

    Hospital safety culture is an integral part of providing high quality care for patients, as well as promoting a safe and healthy environment for healthcare workers. In this article, we explore the extent to which cultural humility, which involves openness to cultural diverse individuals and groups, is related to hospital safety culture. A sample of 2011 hospital employees from four hospitals completed measures of organizational cultural humility and hospital safety culture. Higher perceptions of organizational cultural humility were associated with higher levels of general perceptions of hospital safety, as well as more positive ratings on non-punitive response to error (i.e., mistakes of staff are not held against them), handoffs and transitions, and organizational learning. The cultural humility of one's organization may be an important factor to help improve hospital safety culture. We conclude by discussing potential directions for future research.

  7. 2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

  8. 2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

  9. The Department of Energy nuclear criticality safety program

    International Nuclear Information System (INIS)

    Felty, J.R.

    2004-01-01

    This paper broadly covers key events and activities from which the Department of Energy Nuclear Criticality Safety Program (NCSP) evolved. The NCSP maintains fundamental infrastructure that supports operational criticality safety programs. This infrastructure includes continued development and maintenance of key calculational tools, differential and integral data measurements, benchmark compilation, development of training resources, hands-on training, and web-based systems to enhance information preservation and dissemination. The NCSP was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 97-2, Criticality Safety, and evolved from a predecessor program, the Nuclear Criticality Predictability Program, that was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 93-2, The Need for Critical Experiment Capability. This paper also discusses the role Dr. Sol Pearlstein played in helping the Department of Energy lay the foundation for a robust and enduring criticality safety infrastructure.

  10. 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

  11. 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

  12. Discussion on safety culture general contract model of consultation enterprises

    International Nuclear Information System (INIS)

    Dong Huimin; Zhang Hao

    2012-01-01

    With a high safety requirement, long construction period, a large amount of investment and many influencing factors of the preparation and implementation of project schedule, local nuclear power always is built through EPC. Safety level depends on EPC. Some measures should be taken for local consultation enterprises to improve situation of safety. Some suggestion as follows: safety culture should be received enough attention; management system should be established in according with requirement of safety culture; try to encourage employee involvement; to assess it in time; safety system should be entirely compatible with enterprises system. (authors)

  13. Impact of proof test interval and coverage on probability of failure of safety instrumented function

    International Nuclear Information System (INIS)

    Jin, Jianghong; Pang, Lei; Hu, Bin; Wang, Xiaodong

    2016-01-01

    Highlights: • Introduction of proof test coverage makes the calculation of the probability of failure for SIF more accurate. • The probability of failure undetected by proof test is independently defined as P TIF and calculated. • P TIF is quantified using reliability block diagram and simple formula of PFD avg . • Improving proof test coverage and adopting reasonable test period can reduce the probability of failure for SIF. - Abstract: Imperfection of proof test can result in the safety function failure of safety instrumented system (SIS) at any time in its life period. IEC61508 and other references ignored or only elementarily analyzed the imperfection of proof test. In order to further study the impact of the imperfection of proof test on the probability of failure for safety instrumented function (SIF), the necessity of proof test and influence of its imperfection on system performance was first analyzed theoretically. The probability of failure for safety instrumented function resulted from the imperfection of proof test was defined as probability of test independent failures (P TIF ), and P TIF was separately calculated by introducing proof test coverage and adopting reliability block diagram, with reference to the simplified calculation formula of average probability of failure on demand (PFD avg ). Research results show that: the shorter proof test period and the higher proof test coverage indicate the smaller probability of failure for safety instrumented function. The probability of failure for safety instrumented function which is calculated by introducing proof test coverage will be more accurate.

  14. 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

  15. 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)

  16. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  17. Safety coaches in radiology: decreasing human error and minimizing patient harm

    Energy Technology Data Exchange (ETDEWEB)

    Dickerson, Julie M.; Adams, Janet M. [Cincinnati Children' s Hospital Medical Center, Department of Radiology, MLC 5031, Cincinnati, OH (United States); Koch, Bernadette L.; Donnelly, Lane F. [Cincinnati Children' s Hospital Medical Center, Department of Radiology, MLC 5031, Cincinnati, OH (United States); Cincinnati Children' s Hospital Medical Center, Department of Pediatrics, Cincinnati, OH (United States); Goodfriend, Martha A. [Cincinnati Children' s Hospital Medical Center, Department of Quality Improvement, Cincinnati, OH (United States)

    2010-09-15

    Successful programs to improve patient safety require a component aimed at improving safety culture and environment, resulting in a reduced number of human errors that could lead to patient harm. Safety coaching provides peer accountability. It involves observing for safety behaviors and use of error prevention techniques and provides immediate feedback. For more than a decade, behavior-based safety coaching has been a successful strategy for reducing error within the context of occupational safety in industry. We describe the use of safety coaches in radiology. Safety coaches are an important component of our comprehensive patient safety program. (orig.)

  18. Safety coaches in radiology: decreasing human error and minimizing patient harm

    International Nuclear Information System (INIS)

    Dickerson, Julie M.; Adams, Janet M.; Koch, Bernadette L.; Donnelly, Lane F.; Goodfriend, Martha A.

    2010-01-01

    Successful programs to improve patient safety require a component aimed at improving safety culture and environment, resulting in a reduced number of human errors that could lead to patient harm. Safety coaching provides peer accountability. It involves observing for safety behaviors and use of error prevention techniques and provides immediate feedback. For more than a decade, behavior-based safety coaching has been a successful strategy for reducing error within the context of occupational safety in industry. We describe the use of safety coaches in radiology. Safety coaches are an important component of our comprehensive patient safety program. (orig.)

  19. Safety coaches in radiology: decreasing human error and minimizing patient harm.

    Science.gov (United States)

    Dickerson, Julie M; Koch, Bernadette L; Adams, Janet M; Goodfriend, Martha A; Donnelly, Lane F

    2010-09-01

    Successful programs to improve patient safety require a component aimed at improving safety culture and environment, resulting in a reduced number of human errors that could lead to patient harm. Safety coaching provides peer accountability. It involves observing for safety behaviors and use of error prevention techniques and provides immediate feedback. For more than a decade, behavior-based safety coaching has been a successful strategy for reducing error within the context of occupational safety in industry. We describe the use of safety coaches in radiology. Safety coaches are an important component of our comprehensive patient safety program.

  20. Accounting for Health and Safety costs

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.

    A part of the emerging sustainability management accounting is corporate health and safety performance. One performance dimension is the costs of occupational accidents in companies. The underlying logic for calculating these costs is that if occupational accidents are prevented then these costs...... could be avoided. This chapter presents and discusses selected methods for calculating the costs of occupational accidents. The focus is on presenting the characteristics of each method and disclosing the benefits and drawbacks of each method...

  1. Spanish Nuclear Safety Research under International Frameworks

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Reventos, F.; Ahnert, C.; Jimenez, G.; Queral, C.; Verdu, G.; Miro, R.; Gallardo, S.

    2013-10-01

    The Nuclear Safety research requires a wide international collaboration of several involved groups. In this sense this paper pretends to show several examples of the Nuclear Safety research under international frameworks that is being performed in different Universities and Research Institutions like CIEMAT, Universitat Politecnica de Catalunya (UPC), Universidad Politecnica de Madrid (UPM) and Universitat Politenica de Valencia (UPV). (Author)

  2. Noise in position measurement by centroid calculation

    International Nuclear Information System (INIS)

    Volkov, P.

    1996-01-01

    The position of a particle trajectory in a gaseous (or semiconductor) detector can be measured by calculating the centroid of the induced charge on the cathode plane. The charge amplifiers attached to each cathode strip introduce noise which is added to the signal. This noise broadens the position resolution line. Our article gives an analytical tool to estimate the resolution broadening due to the noise per strip and the number of strips involved in the centroid calculation. It is shown that the position resolution increases faster than the square root of the number of strips involved. We also consider the consequence of added interstrip capacitors, intended to diminish the differential nonlinearity. It is shown that the position error increases slower than linearly with the interstrip capacities, due to the cancellation of correlated noise. The estimation we give, can be applied to calculations of position broadening other than the centroid finding. (orig.)

  3. Using game technologies to improve the safety of construction plant operations.

    Science.gov (United States)

    Guo, Hongling; Li, Heng; Chan, Greg; Skitmore, Martin

    2012-09-01

    Many accidents occur world-wide in the use of construction plant and equipment, and safety training is considered by many to be one of the best approaches to their prevention. However, current safety training methods/tools are unable to provide trainees with the hands-on practice needed. Game technology-based safety training platforms have the potential to overcome this problem in a virtual environment. One such platform is described in this paper - its characteristics are analysed and its possible contribution to safety training identified. This is developed and tested by means of a case study involving three major pieces of construction plant, which successfully demonstrates that the platform can improve the process and performance of the safety training involved in their operation. This research not only presents a new and useful solution to the safety training of construction operations, but illustrates the potential use of advanced technologies in solving construction industry problems in general. Copyright © 2011 Elsevier Ltd. All rights reserved.

  4. System and safety studies of accelerator driven systems for transmutation. Annual report 2007

    International Nuclear Information System (INIS)

    Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang

    2008-05-01

    Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code

  5. State-of-the-art for multiconfiguration Dirac-Fock calculations

    International Nuclear Information System (INIS)

    Desclaux, J.P.

    1981-01-01

    The approximations involved in almost all relativistic calculations are analyzed and one of the most advanced methods, the multiconfiguration Dirac-Fock (MCDF) one, available to carry out high quality atomic calculations for bound states is discussed

  6. HTR-PROTEUS benchmark calculations. Pt. 1. Unit cell results LEUPRO-1 and LEUPRO-2

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Klippel, H.T.; Kuijper, J.C.

    1995-09-01

    In the framework of the IAEA Co-ordinated Research Programme (CRP) on 'Validation of Safety Related Physics Calculations for Low-Enriched (LEU) HTGRs' calculational benchmarks are performed on the basis of LEU-HTR pebble-bed critical experiments carried out in the PROTEUS facility at PSI, Switzerland. Of special interest is the treatment of the double heterogeneity of the fuel and the spherical fuel elements of these pebble bed core configurations. Also of interest is the proper calculation of the safety related physics parameters like the effect of water ingress and control rod worth. This document describes the ECN results of the LEUPRO-1 and LEUPRO-2 unitcell calculations performed with the codes WIMS-E, SCALE-4 and MCNP4A. Results of the LEUPRO-1 unit cell with 20% water ingress in the void is also reported for both the single and the double heterogeneous case. Emphasis is put on the intercomparison of the results obtained by the deterministic codes WIMS-E and SCALE-4, and the Monte Carlo code MCNP4A. The LEUPRO whole core calculations will be reported later. (orig.)

  7. Development of Calculation Algorithm for ECCS Kinematic Shock

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung-Chan; Yoon, Duk-Joo; Ha, Sang-Jun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    The void fraction of inverted U-pipes in front of SI(Safety Injection) pumps impact on the pipe system of ECCS(Emergency Core Cooling Systems). This phenomena is called as 'Kinematic Shock'. The purpose of this paper is to achieve the more exactly calculation when the kinematic shock is calculated by simplified equation. The behavior of the void packet of the ECCS pipes is illustrated by the simplified (other name is kinematic shock equation).. The kinematic shock is defined as the depth of total length of void clusters in the pipes of ECCS when the void cluster is continually reached along the part of pipes in vertical direction. In this paper, the simplified equation is evaluated by comparing calculation error each other.]. The more exact methods of calculating the depth of the kinematic shock in ECCS is achieved. The error of kinematic shock calculation is strongly depended on the calculation search gap and the order of Taylor's expansion. From this study, to select the suitable search gap and the suitable calculation order, differential root method, secant method, and Taylor's expansion form are compared one another.

  8. Towards a Usability and Error "Safety Net": A Multi-Phased Multi-Method Approach to Ensuring System Usability and Safety.

    Science.gov (United States)

    Kushniruk, Andre; Senathirajah, Yalini; Borycki, Elizabeth

    2017-01-01

    The usability and safety of health information systems have become major issues in the design and implementation of useful healthcare IT. In this paper we describe a multi-phased multi-method approach to integrating usability engineering methods into system testing to ensure both usability and safety of healthcare IT upon widespread deployment. The approach involves usability testing followed by clinical simulation (conducted in-situ) and "near-live" recording of user interactions with systems. At key stages in this process, usability problems are identified and rectified forming a usability and technology-induced error "safety net" that catches different types of usability and safety problems prior to releasing systems widely in healthcare settings.

  9. Probabilistic safety analysis using microcomputer

    International Nuclear Information System (INIS)

    Futuro Filho, F.L.F.; Mendes, J.E.S.; Santos, M.J.P. dos

    1990-01-01

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

  10. 78 FR 15747 - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving...

    Science.gov (United States)

    2013-03-12

    ... requestor/petitioner to relief. A requestor/ petitioner who fails to satisfy these requirements with respect..., probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the... assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based...

  11. 78 FR 78402 - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving...

    Science.gov (United States)

    2013-12-26

    ... consideration. The contention must be one which, if proven, would entitle the requestor/petitioner to relief. A... assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based... evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to...

  12. Strength Calculation of Inclined Sections of Reinforced Concrete Elements under Transverse Bending

    Science.gov (United States)

    Filatov, V. B.

    2017-11-01

    The authors propose a design model to determine the strength of inclined sections of bent reinforced concrete elements without shear reinforcement for the action of transverse force taking into account the aggregate interlock forces in the inclined crack. The calculated dependences to find out the components of forces acting in an inclined section are presented. The calculated dependences are obtained from the consideration of equilibrium conditions of the block over the inclined crack. A comparative analysis of the experimental values of the failure loads of the inclined section and the theoretical values obtained for the proposed dependencies and normative calculation methods is performed. It is shown that the proposed design model makes it possible to take into account the effect the longitudinal reinforcement percentage has on the inclined section strength, the element cross section height without the introduction of empirical coefficients which contributes to an increase in the structural safety of design solutions including the safety of high-strength concrete elements.

  13. SGHWR fuel performance, safety and reliability

    International Nuclear Information System (INIS)

    Pickman, D.O.; Inglis, G.H.

    1977-05-01

    The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

  14. Task Group on Safety Margins Action Plan (SMAP). Safety Margins Action Plan - Final Report

    International Nuclear Information System (INIS)

    Hrehor, Miroslav; Gavrilas, Mirela; Belac, Josef; Sairanen, Risto; Bruna, Giovanni; Reocreux, Michel; Touboul, Francoise; Krzykacz-Hausmann, B.; Park, Jong Seuk; Prosek, Andrej; Hortal, Javier; Sandervaag, Odbjoern; Zimmerman, Martin

    2007-01-01

    . Chapter 3 looks at techniques for the deterministic calculation of safety margins and discusses the complementary probabilistic risk assessment techniques needed to generalize safety margins beyond design basis accidents. Chapter 4 examines the definition of safety margin, which is noted to take different meanings in different fields. For example, in civil engineering and applications that deal with the load-strength interference concept, safety margin describes the distance between the means of the load and strength probability density functions with regard to the standard deviation in both. However, in the nuclear industry, the term safety margin evolved to describe the goal of assuring the existence of adequate safety margin in deterministic calculations. Specifically, safety margin refers to keeping the value of a given safety variable under a pre-established safety limit in design basis accidents. Implicitly, safety margin in the nuclear industry is the distance from the safety limit to onset of damage. The SMAP task group fulfilled its first objective by adopting a methodology for quantifying safety margins that merges the deterministic and probabilistic approaches. The methodology described in Chapter 5 is consistent with the definition of safety margin commonly used in the nuclear industry. The metrics of this methodology quantify the change in safety over a range of accident sequences that extend beyond the design bases. However, the methodology is not described in this report to a level that would meet guidance document requirements. This is in part because methods and techniques needed to quantify safety margins in a global manner are evolving, and thus specific guidance rendered at this time would shortly become obsolete. This report presents the framework in sufficient detail to serve as the basis of an analysis and, thus, this report meets the second objective established for the SMAP group. A proof-of-concept application to further aid potential applicants

  15. Legal and governmental infrastructure for nuclear, radiation, radioactive waste and transport safety. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    This publication establishes requirements for legal and governmental responsibilities in respect of the safety of nuclear facilities, the safe use of sources of ionizing radiation, radiation protection, the safe management of radioactive waste and the safe transport of radioactive material. Thus, it covers development of the legal framework for establishing a regulatory body and other actions to achieve effective regulatory control of facilities and activities. Other responsibilities are also covered, such as those for developing the necessary support for safety, involvement in securing third party liability and emergency preparedness

  16. Legal and governmental infrastructure for nuclear, radiation, radioactive waste and transport safety. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    This publication establishes requirements for legal and governmental responsibilities in respect of the safety of nuclear facilities, the safe use of sources of ionizing radiation, radiation protection, the safe management of radioactive waste and the safe transport of radioactive material. Thus, it covers development of the legal framework for establishing a regulatory body and other actions to achieve effective regulatory control of facilities and activities. Other responsibilities are also covered, such as those for developing the necessary support for safety, involvement in securing third party liability and emergency preparedness

  17. Identification of cellular targets involved in cardiac failure caused by PKI in oncology: an approach combining pharmacovigilance and pharmacodynamics.

    Science.gov (United States)

    Patras de Campaigno, Emilie; Bondon-Guitton, Emmanuelle; Laurent, Guy; Montastruc, Francois; Montastruc, Jean-Louis; Lapeyre-Mestre, Maryse; Despas, Fabien

    2017-07-01

    The aims of the present study were to evaluate the risk of cardiac failure (CF) associated with 15 anticancer protein kinase inhibitors (PKIs) through a case/noncase analysis and to identify which PK(s) and pathways are involved in PKI-induced CF. In order to evaluate the risk of CF, adjusted reporting odds ratios (aRORs) were calculated for the 15 anticancer PKIs in the World Health Organization safety report database (VigiBase®). We realised a literature review to identify 21 protein kinases (PKs) that were possibly involved in CF caused by PKIs. Pearson correlation coefficients (r) between aRORs and affinity data of the 15 PKIs for the 21 PKs were calculated to identify the cellular target most likely to be involved in PKI-induced CF. A total of 141 601 individual case safety reports (ICSRs) were extracted from VigiBase® for the following PKIs: afatinib, axitinib, bosutinib, crizotinib, dasatinib, erlotinib, gefitinib, imatinib, lapatinib, nilotinib, pazopanib, ruxolitinib, sorafenib, sunitinib and vandetanib. Among them, 2594 ICSRs concerned CF. The disproportionality analysis revealed that, for dasatinib, imatinib, bosutinib, sunitinib and nilotinib, disproportionality for CF was significantly higher than for other PKIs, with aRORs of 2.52 [95% CI 2.26, 2.82], 1.79 (95% CI 1.57, 2.03), 1.73 (95% CI 1.18, 2.54), 1.67 (95% CI 1.51, 1.84) and 1.38 (95% CI 1.18, 1.61), respectively. Significant values for correlation coefficients between the product of dissociation constant (pKd) and aROR were observed for two non-receptor protein kinases: ABL1 (non-phosphorylated and phosphorylated forms) and ABL2 protein kinases, with values of r = 0.83 (P = 0.0001), r = 0.75 (P = 0.0014) and r = 0.78 (P = 0.0006), respectively. We observed a higher disproportionality for CF with dasatinib, imatinib, bosutinib, sunitinib and nilotinib than with other PKIs. In addition, the study highlighted the role of ABL tyrosine kinases in CF caused by anticancer PKIs. © 2017 The British

  18. Relationship between organisational safety culture dimensions and crashes.

    Science.gov (United States)

    Varmazyar, Sakineh; Mortazavi, Seyed Bagher; Arghami, Shirazeh; Hajizadeh, Ebrahim

    2016-01-01

    Knowing about organisational safety culture in public transportation system can provide an appropriate guide to establish effective safety measures and interventions to improve safety at work. The aim of this study was investigation of association between safety culture dimensions (leadership styles and company values, usage of crashes information and prevention programmes, management commitment and safety policy, participation and control) with involved self-reported crashes. The associations were considered through Spearman correlation, Pearson chi-square test and logistic regression. The results showed an association among self-reported crashes (occurrence or non-occurrence) and factors including leadership styles and company values; management commitment and safety policy; and control. Moreover, it was found a negative correlation and an odds ratio less than one between control and self-reported crashes.

  19. Managing patient safety through NPSGs and employee performance.

    Science.gov (United States)

    Adair, Liberty

    2010-01-01

    Patient safety can only exist in a culture of patient safety, which implies it is a value perceived by all. Culture predicts safety outcomes and leadership predicts the culture. Leaders are obligated to continually mitigate hazard and take action consciously. Healthcare workers should focus on preventing and reporting mistakes with the National Patient Safety Goals (NPSGs) in mind. These include: accuracy of patient identification, effectiveness of communication among caregivers, improving safety of medications, reducing infections, reducing risk of falls, and encouraging patients to be involved in care. Poor performers and reckless behavior need to be mitigated. If employees recognize their roles in the process, feel empowered,and have appropriate tools, resources,and data to implement solutions, errors can be avoided and patient safety becomes paramount.

  20. 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)

  1. Code of safety for nuclear merchant ships

    International Nuclear Information System (INIS)

    1982-01-01

    The Code is in chapters, entitled: general (including general safety principles and principles of risk acceptance); design criteria and conditions; ship design, construction and equipment; nuclear steam supply system; machinery and electrical installations; radiation safety (including radiological protection design; protection of persons; dosimetry; radioactive waste management); operation (including emergency operation procedures); surveys. Appendices cover: sinking velocity calculations; seaway loads depending on service periods; safety assessment; limiting dose-equivalent rates for different areas and spaces; quality assurance programme; application of single failure criterion. Initial application of the Code is restricted to conventional types of ships propelled by nuclear propulsion plants with pressurized light water type reactors. (U.K.)

  2. SCALE criticality safety verification and validation package

    International Nuclear Information System (INIS)

    Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

    1998-01-01

    Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

  3. AEC controlled area safety program

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, D W [Nevada Operations Office, Atomic Energy Commission, Las Vegas, NV (United States)

    1969-07-01

    The detonation of underground nuclear explosives and the subsequent data recovery efforts require a comprehensive pre- and post-detonation safety program for workers within the controlled area. The general personnel monitoring and environmental surveillance program at the Nevada Test Site are presented. Some of the more unusual health-physics aspects involved in the operation of this program are also discussed. The application of experience gained at the Nevada Test Site is illustrated by description of the on-site operational and safety programs established for Project Gasbuggy. (author)

  4. AEC controlled area safety program

    International Nuclear Information System (INIS)

    Hendricks, D.W.

    1969-01-01

    The detonation of underground nuclear explosives and the subsequent data recovery efforts require a comprehensive pre- and post-detonation safety program for workers within the controlled area. The general personnel monitoring and environmental surveillance program at the Nevada Test Site are presented. Some of the more unusual health-physics aspects involved in the operation of this program are also discussed. The application of experience gained at the Nevada Test Site is illustrated by description of the on-site operational and safety programs established for Project Gasbuggy. (author)

  5. Assessing medical students' perceptions of patient safety: the medical student safety attitudes and professionalism survey.

    Science.gov (United States)

    Liao, Joshua M; Etchegaray, Jason M; Williams, S Tyler; Berger, David H; Bell, Sigall K; Thomas, Eric J

    2014-02-01

    To develop and test the psychometric properties of a survey to measure students' perceptions about patient safety as observed on clinical rotations. In 2012, the authors surveyed 367 graduating fourth-year medical students at three U.S. MD-granting medical schools. They assessed the survey's reliability and construct and concurrent validity. They examined correlations between students' perceptions of organizational cultural factors, organizational patient safety measures, and students' intended safety behaviors. They also calculated percent positive scores for cultural factors. Two hundred twenty-eight students (62%) responded. Analyses identified five cultural factors (teamwork culture, safety culture, error disclosure culture, experiences with professionalism, and comfort expressing professional concerns) that had construct validity, concurrent validity, and good reliability (Cronbach alphas > 0.70). Across schools, percent positive scores for safety culture ranged from 28% (95% confidence interval [CI], 13%-43%) to 64% (30%-98%), while those for teamwork culture ranged from 47% (32%-62%) to 74% (66%-81%). They were low for error disclosure culture (range: 10% [0%-20%] to 27% [20%-35%]), experiences with professionalism (range: 7% [0%-15%] to 23% [16%-30%]), and comfort expressing professional concerns (range: 17% [5%-29%] to 38% [8%-69%]). Each cultural factor correlated positively with perceptions of overall patient safety as observed in clinical rotations (r = 0.37-0.69, P safety behavioral intent item. This study provided initial evidence for the survey's reliability and validity and illustrated its applicability for determining whether students' clinical experiences exemplify positive patient safety environments.

  6. AER Working Group D on VVER safety analysis minutes of the meeting in Rez, Czech Republic 18-20 May 1998

    International Nuclear Information System (INIS)

    Siltanen, P.

    1998-01-01

    AER Working Group D on VVER reactor safety analysis held its seventh meeting in Hotel Vltava in Rez near Prague during the period 18-20 May 1998. There were altogether 11 participants from 8 member organisations. The coordinator for the working group, Mr. P. Siltanen (IVO) served as chairman. In addition to the general information exchange on recent activities, the topics of the meeting included: First review of solutions to the 3-dimensional AER Dynamic Benchmark Problem No. 5 on a steam line break accident. This benchmark involves a break of the main steam header. Safety analysis of reactivity events. Recent code development work and fuel behaviour. Coolant mixing calculations and experiments related to diluted slugs. A list of participants and a list of handouts distributed at the meeting are attached to the minutes. (author)

  7. The relationship between patient safety climate and occupational safety climate in healthcare - A multi-level investigation.

    Science.gov (United States)

    Pousette, Anders; Larsman, Pernilla; Eklöf, Mats; Törner, Marianne

    2017-06-01

    Patient safety climate/culture is attracting increasing research interest, but there is little research on its relation with organizational climates regarding other target domains. The aim of this study was to investigate the relationship between patient safety climate and occupational safety climate in healthcare. The climates were assessed using two questionnaires: Hospital Survey on Patient Safety Culture and Nordic Occupational Safety Climate Questionnaire. The final sample consisted of 1154 nurses, 886 assistant nurses, and 324 physicians, organized in 150 work units, within hospitals (117units), primary healthcare (5units) and elderly care (28units) in western Sweden, which represented 56% of the original sample contacted. Within each type of safety climate, two global dimensions were confirmed in a higher order factor analysis; one with an external focus relative the own unit, and one with an internal focus. Two methods were used to estimate the covariation between the global climate dimensions, in order to minimize the influence of bias from common method variance. First multilevel analysis was used for partitioning variances and covariances in a within unit part (individual level) and a between unit part (unit level). Second, a split sample technique was used to calculate unit level correlations based on aggregated observations from different respondents. Both methods showed associations similar in strength between the patient safety climate and the occupational safety climate domains. The results indicated that patient safety climate and occupational safety climate are strongly positively related at the unit level, and that the same organizational processes may be important for the development of both types of organizational climate. Safety improvement interventions should not be separated in different organizational processes, but be planned so that both patient safety and staff safety are considered concomitantly. Copyright © 2017 National Safety

  8. Source term calculations - Ringhals 2 PWR

    International Nuclear Information System (INIS)

    Johansson, L.L.

    1998-02-01

    This project was performed within the fifth and final phase of sub-project RAK-2.1 of the Nordic Co-operative Reactor Safety Program, NKS.RAK-2.1 has also included studies of reflooding of degraded core, recriticality and late phase melt progression. Earlier source term calculations for Swedish nuclear power plants are based on the integral code MAAP. A need was recognised to compare these calculations with calculations done with mechanistic codes. In the present work SCDAP/RELAP5 and CONTAIN were used. Only limited results could be obtained within the frame of RAK-2.1, since many problems were encountered using the SCDAP/RELAP5 code. The main obstacle was the extremely long execution times of the MOD3.1 version, but also some dubious fission product calculations. However, some interesting results were obtained for the studied sequence, a total loss of AC power. The report describes the modelling approach for SCDAP/RELAP5 and CONTAIN, and discusses results for the transient including the event of a surge line creep rupture. The study will probably be completed later, providing that an improved SCDAP/RELAP5 code version becomes available. (au) becomes available. (au)

  9. 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

  10. Health and safety education for joint occupational health and safety committees

    Directory of Open Access Journals (Sweden)

    Myriam Mahecha Angulo

    2015-09-01

    Full Text Available Objective: To build a proposal to develop the educational process in health and safety joint committees aimed at safety and health at work (copasst. Methodology: Qualitative, descriptive study in which an in-depth interview to 32 copasst assets was made. Each interview was transcribed and interpreted by applying check with participants, finding meaningful statements, organizing groups of subjects, exhaustive description and validation with participants. The information was placed in the categories planning, organization, development, evaluation and feedback, emerging the following categories: responsible for processes management; planning, place and frequency of educational sessions; topics; format of sessions; involving/ development of sessions; understanding of the issues; applicability to daily life and work environment; applicability to personal/professional life and to the organization. Results: From emerging categories and according to the conceptual framework on adult health education and health and safety for workers, a participatory methodology for the development of educational processes with copasst was built. Conclusions: According to the statement by the members of the copasst, educational processes in health and safety, as they are developed at present, preclude them from achieving necessary competences to perform its functions, thus they are irrelevant.

  11. 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)

  12. Minimum qualifications for nuclear criticality safety professionals

    International Nuclear Information System (INIS)

    Ketzlach, N.

    1990-01-01

    A Nuclear Criticality Technology and Safety Training Committee has been established within the U.S. Department of Energy (DOE) Nuclear Criticality Safety and Technology Project to review and, if necessary, develop standards for the training of personnel involved in nuclear criticality safety (NCS). The committee is exploring the need for developing a standard or other mechanism for establishing minimum qualifications for NCS professionals. The development of standards and regulatory guides for nuclear power plant personnel may serve as a guide in developing the minimum qualifications for NCS professionals

  13. Calculation of Environmental Conditions in NEK Intermediate Building Following HELB

    International Nuclear Information System (INIS)

    Grgic, D.; Spalj, S.; Basic, I.

    1998-01-01

    The purpose of Equipment Qualification (EQ) in nuclear power plants is to ensure the capability of safety related equipment to perform its function on demand under postulated service conditions, including harsh accident environment (e.g. Loss of Coolant Accident - LOCA, High Energy Line Break - HELB). The determination of the EQ conditions and zones is one of the basic steps in the frame of the overall EQ project. The EQ parameters (temperature, pressure, relative humidity, chemical spray, submergence, radiation) should be defined for all locations of the plant containing equipment important to safety. This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside Intermediate Building (IB) of Krsko NPP after the postulated HELB. The RELAP5/mod2 computer code was used for the determination of HELB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko IB. (author)

  14. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  15. [Concept analysis of a participatory approach to occupational safety and health].

    Science.gov (United States)

    Yoshikawa, Etsuko

    2013-01-01

    The purpose of this study was to analyze a participatory approach to occupational safety and health, and to examine the possibility of applying the concept to the practice and research of occupational safety and health. According to Rodger's method, descriptive data concerning antecedents, attributes and consequences were qualitatively analyzed. A total of 39 articles were selected for analysis. Attributes with a participatory approach were: "active involvement of both workers and employers", "focusing on action-oriented low-cost and multiple area improvements based on good practices", "the process of emphasis on consensus building", and "utilization of a local network". Antecedents of the participatory approach were classified as: "existing risks at the workplace", "difficulty of occupational safety and health activities", "characteristics of the workplace and workers", and "needs for the workplace". The derived consequences were: "promoting occupational safety and health activities", "emphasis of self-management", "creation of safety and healthy workplace", and "contributing to promotion of quality of life and productivity". A participatory approach in occupational safety and health is defined as, the process of emphasis on consensus building to promote occupational safety and health activities with emphasis on self-management, which focuses on action-oriented low-cost and multiple area improvements based on good practices with active involvement of both workers and employers through utilization of local networks. We recommend that the role of the occupational health professional be clarified and an evaluation framework be established for the participatory approach to promote occupational safety and health activities by involving both workers and employers.

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1983-02-01

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

  18. A Study on the Estimation Method of Risk Based Area for Jetty Safety Monitoring

    Directory of Open Access Journals (Sweden)

    Byeong-Wook Nam

    2015-09-01

    Full Text Available Recently, the importance of safety-monitoring systems was highlighted by the unprecedented collision between a ship and a jetty in Yeosu. Accordingly, in this study, we introduce the concept of risk based area and develop a methodology for a jetty safety-monitoring system. By calculating the risk based areas for a ship and a jetty, the risk of collision was evaluated. To calculate the risk based areas, we employed an automatic identification system for the ship, stopping-distance equations, and the regulation velocity near the jetty. In this paper, we suggest a risk calculation method for jetty safety monitoring that can determine the collision probability in real time and predict collisions using the amount of overlap between the two calculated risk based areas. A test was conducted at a jetty control center at GS Caltex, and the effectiveness of the proposed risk calculation method was verified. The method is currently applied to the jetty-monitoring system at GS Caltex in Yeosu for the prevention of collisions.

  19. Qualification of calculation aids for PSA

    International Nuclear Information System (INIS)

    Goetz, K.; Hennigs, W.; Kirstein, B.M.; Reinhardt, C.

    1998-01-01

    In Germany Probabilistic Safety Analysis (PSA) are part of the evaluation of a nuclear power plants safety. The German PSA guide stipulates that the used software must enable a PSA according to the state of the art. This software must be qualified to assure that its features, mathematic methods and its performance enable a PSA like this. In this research work specifications and requirements are developed, which allow the testing of software. A procedure was developed to qualify PSA software according to the PSA guide and the experiences of users of PSA. Setting up a procedure, a tool for a systematic and uniform examination was crated. Additionally the options, mathematic fundamentals and performance of PSA-programs were analyzed. According to this all programs that were analyzed are capable to sovle their original task, that is the calculation of the safety of high available system based on high available components. Against that the requirements of modern PSA, e.g. to handle less available functions, HRA and fire analyses, based on the use of modern software and the implementation of new developments in the field of PSA are not supported adequately by all programs. (orig.) [de

  20. Feasibility study and uncertainties in the validation of an existing safety-related control circuit with the ISO 13849-1:2006 design standard

    International Nuclear Information System (INIS)

    Jocelyn, Sabrina; Baudoin, James; Chinniah, Yuvin; Charpentier, Philippe

    2014-01-01

    In industry, machine users and people who modify or integrate equipment often have to evaluate the safety level of a safety-related control circuit that they have not necessarily designed. The modifications or integrations may involve work to make an existing machine that does not comply with normative or regulatory specifications safe. However, how can a circuit performing a safety function be validated a posteriori? Is the validation exercise feasible? What are the difficulties and limitations of such a procedure? The aim of this article is to answer these questions by presenting a validation study of a safety function of an existing machine. A plastic injection molding machine is used for this study, as well as standard ISO 13849-1:2006. Validation consists of performing an a posteriori (post-design) estimation of the performance level of the safety function. The procedure is studied for two contexts of use of the machine: in industry, and in laboratory. The calculations required by the ISO standard were done using Excel, followed by SIStema software. It is shown that, based on the context of use, the estimated performance level was different for the same safety-related circuit. The variability in the results is explained by the assumptions made by the person undertaking the validation without the involvement of the machine designer. - Highlights: • Validation of the performance level of a safety function is undertaken. • An injection molding machine and ISO 13849-1:2006 standard are used for the procedure. • The procedure is undertaken for two contexts of use of the machine. • In this study, the performance level depends on the context of use. • The assumptions made throughout the study partially explain this difference