WorldWideScience

Sample records for preliminary documented safety

  1. Preliminary safety information document for the standard MHTGR. Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-01-01

    This report contains information concerning: operational radionuclide control; occupational radiation protection, conduct of operations; initial test program; safety analysis; technical specifications; and quality assurance. (JDB)

  2. Preliminary Safety Information Document for the Standard MHTGR. Volume 1, (includes latest Amendments)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-01-01

    With NRC concurrence, the Licensing Plan for the Standard HTGR describes an application program consistent with 10CFR50, Appendix O to support a US Nuclear Regulatory Commission (NRC) review and design certification of an advanced Standard modular High Temperature Gas-Cooled Reactor (MHTGR) design. Consistent with the NRC's Advanced Reactor Policy, the Plan also outlines a series of preapplication activities which have as an objective the early issuance of an NRC Licensability Statement on the Standard MHTGR conceptual design. This Preliminary Safety Information Document (PSID) has been prepared as one of the submittals to the NRC by the US Department of Energy in support of preapplication activities on the Standard MHTGR. Other submittals to be provided include a Probabilistic Risk Assessment, a Regulatory Technology Development Plan, and an Emergency Planning Bases Report.

  3. Preliminary safety evaluation, based on initial site investigation data. Planning document

    International Nuclear Information System (INIS)

    Hedin, Allan

    2002-12-01

    This report is a planning document for the preliminary safety evaluations (PSE) to be carried out at the end of the initial stage of SKBs ongoing site investigations for a deep repository for spent nuclear fuel. The main purposes of the evaluations are to determine whether earlier judgements of the suitability of the candidate area for a deep repository with respect to long-term safety holds up in the light of borehole data and to provide feed-back to continued site investigations and site specific repository design. The preliminary safety evaluations will be carried out by a safety assessment group, based on a site model, being part of a site description, provided by a site modelling group and a repository layout within that model suggested by a repository engineering group. The site model contains the geometric features of the site as well as properties of the host rock. Several alternative interpretations of the site data will likely be suggested. Also the biosphere is included in the site model. A first task for the PSE will be to compare the rock properties described in the site model to previously established criteria for a suitable host rock. This report gives an example of such a comparison. In order to provide more detailed feedback, a number of thermal, hydrological, mechanical and chemical analyses of the site will also be included in the evaluation. The selection of analyses is derived from the set of geosphere and biosphere analyses preliminarily planned for the comprehensive safety assessment named SR-SITE, which will be based on a complete site investigation. The selection is dictated primarily by the expected feedback to continued site investigations and by the availability of data after the PSE. The repository engineering group will consider several safety related factors in suggesting a repository layout: Thermal calculations will be made to determine a minimum distance between canisters avoiding canister surface temperatures above 100 deg C

  4. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    International Nuclear Information System (INIS)

    Herborn, D.I.

    1991-10-01

    The requirements for Westinghouse Hanford independent review of the Preliminary Safety Analysis Report (PSAR) are contained in Section 1.0, Subsection 4.3 of WCH-CM-4-46. Specifically, this manual requires the following: (1) Formal functional reviews of the HWVP PSAR by the future operating organization (HWVP Operations), and the independent review organizations (HWVP and Environmental Safety Assurance, Environmental Assurance, and Quality Assurance); and (2) Review and approval of the HWVP PSAR by the Tank Waste Disposal (TWD) Subcouncil of the Safety and Environmental Advisory Council (SEAC), which provides independent advice to the Westinghouse Hanford President and executives on matters of safety and environmental protection. 7 refs

  5. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    International Nuclear Information System (INIS)

    Herborn, D.I.

    1993-11-01

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

  6. Generic safety documentation model

    International Nuclear Information System (INIS)

    Mahn, J.A.

    1994-04-01

    This document is intended to be a resource for preparers of safety documentation for Sandia National Laboratories, New Mexico facilities. It provides standardized discussions of some topics that are generic to most, if not all, Sandia/NM facilities safety documents. The material provides a ''core'' upon which to develop facility-specific safety documentation. The use of the information in this document will reduce the cost of safety document preparation and improve consistency of information

  7. Liquid-metal fast-breeder reactors: Preliminary safety and environmental information document. Volume VI

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning LMFBR design characteristics; uranium-plutonium/uranium recycle homogeneous core; uranium-plutonium/uranium spiked recycle heterogeneous core; uranium-plutonium/uranium spiked recycle homogeneous core; uranium-plutonium/thorium spiked recycle heterogeneous core; uranium-plutonium/thorium spiked recycle homogeneous core; thorium-plutonium/thorium spiked recycle homogeneous core; denatured uranium-233/thorium cycle homogeneous core; safety consideration for the LMFBR; and environmental considerations

  8. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    International Nuclear Information System (INIS)

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

  9. Environment, Health, and Safety - Construction Subcontractors Documents |

    Science.gov (United States)

    NREL Environment, Health, and Safety - Construction Subcontractors Documents Environment Environment, Health and Safety (EH&S) requirements are understood by construction subcontractors and with these requirements before submitting proposals and/or environment, health and safety plans for the

  10. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  11. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    International Nuclear Information System (INIS)

    1980-01-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

  12. PUREX Deactivation Health and Safety documentation

    International Nuclear Information System (INIS)

    Dodd, E.N. III.

    1995-01-01

    The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D ampersand D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety

  13. PUREX Deactivation Health and Safety documentation

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, E.N. III

    1995-01-01

    The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D&D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety.

  14. Using Addenda in Documented Safety Analysis Reports

    International Nuclear Information System (INIS)

    Swanson, D.S.; Thieme, M.A.

    2003-01-01

    This paper discusses the use of addenda to the Radioactive Waste Management Complex (RWMC) Documented Safety Analysis (DSA) located at the Idaho National Engineering and Environmental Laboratory (INEEL). Addenda were prepared for several systems and processes at the facility that lacked adequate descriptive information and hazard analysis in the DSA. They were also prepared for several new activities involving unreviewed safety questions (USQs). Ten addenda to the RWMC DSA have been prepared since the last annual update

  15. Preliminary safety design analysis of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  16. Simplifying documentation while approaching site closure: integrated health and safety plans as documented safety analysis

    International Nuclear Information System (INIS)

    Brown, Tulanda

    2003-01-01

    At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). By isolating any remediation activities that deal with Enriched Restricted Materials, the SBRs and PRs assure that the hazard categories of former nuclear facilities undergoing remediation remain less than Nuclear. These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 150 structures, including six major nuclear production plants. This paper presents the FCP method for maintaining safety basis documentation, using the D and D I-HASP as an example

  17. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  18. FEP catalogue for the VSG. Documentation. Report on the work package 7. Preliminary safety analysis Gorleben (VSG); FEP-Katalog fuer die VSG. Dokumentation. Bericht zum Arbeitspaket 7. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Behlau, Joachim [Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover (Germany); Beuth, Thomas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)] [and others

    2012-06-15

    The report is a compendium of the FEP (features, events, processes) data base that was developed in the frame of the preliminary safety analysis Gorleben (VSG). For each FEP issue the information includes the following subchapters: definition, general information and examples, status at the site, site-specific impacts, temporal restriction, conditional incidence rate, effects on subsidiary systems, adverse effects on the function of the initial barriers, justification, direct dependencies, open questions, references.

  19. Safety performance of preliminary KALIMER conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

    1999-07-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  20. Safety performance of preliminary KALIMER conceptual design

    International Nuclear Information System (INIS)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

    1999-01-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  1. Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

    International Nuclear Information System (INIS)

    Dr. Michael A. Lehto; MAL

    2007-01-01

    A recent accomplishment of the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Nuclear Safety analysis group was to obtain DOE-ID approval for the inter-facility transfer of greater-than-Hazard-Category-3 quantity radioactive/fissionable waste in Department of Transportation (DOT) Type A drums at MFC. This accomplishment supported excellence in operations through safety analysis by better integrating nuclear safety requirements with waste requirements in the Transportation Safety Document (TSD); reducing container and transport costs; and making facility operations more efficient. The MFC TSD governs and controls the inter-facility transfer of greater-than-Hazard-Category-3 radioactive and/or fissionable materials in non-DOT approved containers. Previously, the TSD did not include the capability to transfer payloads of greater-than-Hazard-Category-3 radioactive and/or fissionable materials using DOT Type A drums. Previous practice was to package the waste materials to less-than-Hazard-Category-3 quantities when loading DOT Type A drums for transfer out of facilities to reduce facility waste accumulations. This practice allowed operations to proceed, but resulted in drums being loaded to less than the Waste Isolation Pilot Plant (WIPP) waste acceptance criteria (WAC) waste limits, which was not cost effective or operations friendly. An improved and revised safety analysis was used to gain DOE-ID approval for adding this container configuration to the MFC TSD safety basis. In the process of obtaining approval of the revised safety basis, safety analysis practices were used effectively to directly support excellence in operations. Several factors contributed to the success of MFC's effort to obtain approval for the use of DOT Type A drums, including two practices that could help in future safety basis changes at other facilities. (1) The process of incorporating the DOT Type A drums into the TSD at MFC helped to better integrate nuclear safety

  2. Criteria Document for B-plant's Surveillance and Maintenance Phase Safety Basis Document

    International Nuclear Information System (INIS)

    SCHWEHR, B.A.

    1999-01-01

    This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S and M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S and M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S and M Phase safety basis document

  3. Preliminary safety assessment of the WIPP facility

    International Nuclear Information System (INIS)

    Balestri, R.J.; Torres, B.W.; Pahwa, S.B.; Brannen, J.P.

    1979-01-01

    This paper summarizes the efforts to perform a safety assessment of the Waste Isolation Pilot Plant (WIPP) facility being proposed for southeastern New Mexico. This preliminary safety assessment is limited to a consequence assessment in terms of the dose to a maximally exposed individual as a result of introducing the radionuclides into the biosphere. The extremely low doses to the organs as a result of the liquid breach scenarios are contrasted with the background radiation

  4. Preliminary safety analysis report for the TFTR

    International Nuclear Information System (INIS)

    Lind, K.E.; Levine, J.D.; Howe, H.J.

    A Preliminary Safety Analysis Report has been prepared for the Tokamak Fusion Test Reactor. No accident scenarios have been identified which would result in exposures to on-site personnel or the general public in excess of the guidelines defined for the project by DOE

  5. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    BERGMANN, D.W.

    1999-01-01

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

  6. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

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

  7. Response to Absorber-Focus Coil Preliminary Safety Review Panel

    International Nuclear Information System (INIS)

    Barr, Giles; Baynham, Elwyn; Black, Edgar; Bradshaw, Tom; Cummings, Mary Anne; Green, Michael A.; Ishimoto, Shigeru; Ivanyushenkov, Yury; Lau, Wing; Zisman, Michael S.

    2004-01-01

    In this document we provide responses to the various issues raised in the report of the Preliminary Safety Review Panel (see http://mice.iit.edu/mnp/MICE0069.pdf). In some cases we have made design changes in response to the Panels suggestions. In other cases, we have chosen not to do so. In a few cases, we indicate our plans, although the tasks have not yet been completed. For simplicity, the responses are organized along the same lines as those of the Panel Report

  8. A Generic Software Safety Document Generator

    Science.gov (United States)

    Denney, Ewen; Venkatesan, Ram Prasad

    2004-01-01

    Formal certification is based on the idea that a mathematical proof of some property of a piece of software can be regarded as a certificate of correctness which, in principle, can be subjected to external scrutiny. In practice, however, proofs themselves are unlikely to be of much interest to engineers. Nevertheless, it is possible to use the information obtained from a mathematical analysis of software to produce a detailed textual justification of correctness. In this paper, we describe an approach to generating textual explanations from automatically generated proofs of program safety, where the proofs are of compliance with an explicit safety policy that can be varied. Key to this is tracing proof obligations back to the program, and we describe a tool which implements this to certify code auto-generated by AutoBayes and AutoFilter, program synthesis systems under development at the NASA Ames Research Center. Our approach is a step towards combining formal certification with traditional certification methods.

  9. Preliminary Evaluation of an Aviation Safety Thesaurus' Utility for Enhancing Automated Processing of Incident Reports

    Science.gov (United States)

    Barrientos, Francesca; Castle, Joseph; McIntosh, Dawn; Srivastava, Ashok

    2007-01-01

    This document presents a preliminary evaluation the utility of the FAA Safety Analytics Thesaurus (SAT) utility in enhancing automated document processing applications under development at NASA Ames Research Center (ARC). Current development efforts at ARC are described, including overviews of the statistical machine learning techniques that have been investigated. An analysis of opportunities for applying thesaurus knowledge to improving algorithm performance is then presented.

  10. A graded approach to safety documentation at processing facilities

    International Nuclear Information System (INIS)

    Cowen, M.L.

    1992-01-01

    Westinghouse Savannah River Company (WSRC) has over 40 major Safety Analysis Reports (SARs) in preparation for non-reactor facilities. These facilities include nuclear material production facilities, waste management facilities, support laboratories and environmental remediation facilities. The SARs for these various projects encompass hazard levels from High to Low, and mission times from startup, through operation, to shutdown. All of these efforts are competing for scarce resources, and therefore some mechanism is required for balancing the documentation requirements. Three of the key variables useful for the decision making process are Depth of Safety Analysis, Urgency of Safety Analysis, and Resource Availability. This report discusses safety documentation at processing facilities

  11. Preliminary safety evaluation for CSR1000 with passive safety system

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

    2014-01-01

    Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

  12. Advanced Photon Source experimental beamline Safety Assessment Document: Addendum to the Advanced Photon Source Accelerator Systems Safety Assessment Document (APS-3.2.2.1.0)

    International Nuclear Information System (INIS)

    1995-01-01

    This Safety Assessment Document (SAD) addresses commissioning and operation of the experimental beamlines at the Advanced Photon Source (APS). Purpose of this document is to identify and describe the hazards associated with commissioning and operation of these beamlines and to document the measures taken to minimize these hazards and mitigate the hazard consequences. The potential hazards associated with the commissioning and operation of the APS facility have been identified and analyzed. Physical and administrative controls mitigate identified hazards. No hazard exists in this facility that has not been previously encountered and successfully mitigated in other accelerator and synchrotron radiation research facilities. This document is an updated version of the APS Preliminary Safety Analysis Report (PSAR). During the review of the PSAR in February 1990, the APS was determined to be a Low Hazard Facility. On June 14, 1993, the Acting Director of the Office of Energy Research endorsed the designation of the APS as a Low Hazard Facility, and this Safety Assessment Document supports that designation

  13. Environmental restoration and decontamination and decommissioning safety documentation

    International Nuclear Information System (INIS)

    Hansen, J.L.; Frauenholz, L.H.; Kerr, N.R.

    1993-01-01

    This document presents recommendations of a working group designated by the Environmental Restoration and Remediation (ER) and Decontamination and Decommissioning (D ampersand D) subcommittees of the Westinghouse M ampersand O (Management and Operation) Nuclear Facility Safety Committee. A commonalty of approach to safety documentation specific to ER and D ampersand D activities was developed and is summarized below. Allowance for interpretative tolerance and documentation flexibility appropriate to the activity, graded for hazard category, duration, and complexity, was a primary consideration in development of this guidance

  14. Upgrading safety documentation for exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    In view of the generally small regulatory staffs of importing countries, suggestions are given for upgrading the ''export edition'' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which would highlight the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical operating regulations (Technical Specifications for Operation). (author)

  15. Preparation of safety and regulatory document for BARC Facilities

    International Nuclear Information System (INIS)

    Prasad, S.S.; Jayarajan, K.

    2017-01-01

    In India, the necessary codes and safety guidelines for achieving the safety objectives are provided by the Atomic Energy Regulatory Board (AERB), which are in conformity with the principles of radiation protection as formulated by the International Council of Radiation Protection (ICRP) and International Atomic Energy Agency (IAEA). The same is followed by BARC Safety Council (BSC), which is the regulatory body for the BARC facilities. In addition to all types of fuel cycle facilities, BSC regulates safety of many types of conventional facilities. Many such types of facilities and projects are not under the regulatory purview of AERB. Therefore, the Council has also initiated a programme for development and publication of safety documents for installations in BARC in the fields/ topics yet not addressed by IAEA or AERB. This makes the task pioneering, as some of the areas taken up for defining the regulatory requirements are new, where standard regulatory documents are not available

  16. Preliminary safety analysis of the Gorleben site

    International Nuclear Information System (INIS)

    Bracke, G.; Fischer-Appelt, K.

    2014-01-01

    The safety requirements governing the final disposal of heat-generating radioactive waste in Germany were implemented by the Federal Ministry of Environment, Natural Conservation and Nuclear Safety (BMU) in 2010. The Ministry considers as a fundamental objective the protection of man and the environment against the hazards of radioactive waste. Unreasonable burdens and obligation for future generations shall be avoided. The main safety principles are concentration and inclusion of radioactive and other pollutants in a containment-providing rock zone. Any release of radioactive nuclides may increase the risk for men and the environment only negligibly compared to natural radiation exposure. No intervention or maintenance work shall be necessary in the post-closure phase. Retrieval/recovery of the waste shall be possible up to 500 years after closure. The Gorleben salt dome has been discussed since the 1970's as a possible repository site for heat-generating radioactive waste in Germany. The objective of the project preliminary safety analysis of the Gorleben site (VSG) was to assess if repository concepts at the Gorleben site or other sites with a comparable geology could comply with these requirements based on currently available knowledge (Fischer-Appelt, 2013; Bracke, 2013). In addition to this it was assessed if methodological approaches can be used for a future site selection procedure and which technological and conceptual considerations can be transferred to other geological situations. The objective included the compilation and review of the available exploration data of the Gorleben site and on disposal in salt rock, the development of repository designs, and the identification of the needs for future R and D work and further site investigations. (authors)

  17. Documents pertaining to safety control of nuclear facilities

    International Nuclear Information System (INIS)

    1998-01-01

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls the safety of nuclear facilities in Finland. This control encompasses on one hand the evaluation of plant safety on the basis of plans and analyses pertaining to the plant and on the other hand the inspection of plant structures, systems and components as well as of operational activity. STUK also monitors plants operational experience feedback and technical developments in the field, as well as the development of safety research and takes the necessary measures on their basis. Guide YVL 1.1 describes how STUK controls the design, construction and operation of nuclear power plants. The documents to be submitted to STUK are described in the nuclear energy legislation and YVL guides. This guide presents the mode of delivery, quality, contents and number of documents to be submitted to STUK

  18. Evolution of Safety Basis Documentation for the Fernald Site

    International Nuclear Information System (INIS)

    Brown, T.; Kohler, S.; Fisk, P.; Krach, F.; Klein, B.

    2004-01-01

    The objective of the Department of Energy's (DOE) Fernald Closure Project (FCP), in suburban Cincinnati, Ohio, is to safely complete the environmental restoration of the Fernald site by 2006. Over 200 out of 220 total structures, at this DOE plant site which processed uranium ore concentrates into high-purity uranium metal products, have been safely demolished, including eight of the nine major production plants. Documented Safety Analyses (DSAs) for these facilities have gone through a process of simplification, from individual operating Safety Analysis Reports (SARs) to a single site-wide Authorization Basis containing nuclear facility Bases for Interim Operations (BIOs) to individual project Auditable Safety Records (ASRs). The final stage in DSA simplification consists of project-specific Integrated Health and Safety Plans (I-HASPs) and Nuclear Health and Safety Plans (N-HASPs) that address all aspects of safety, from the worker in the field to the safety basis requirements preserving the facility/activity hazard categorization. This paper addresses the evolution of Safety Basis Documentation (SBD), as DSAs, from production through site closure

  19. Supporting Fernald Site Closure with Integrated Health and Safety Plans as Documented Safety Analyses

    International Nuclear Information System (INIS)

    Kohler, S.; Brown, T.; Fisk, P.; Krach, F.; Klein, B.

    2004-01-01

    At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 200 structures, including eight major nuclear production plants. There is one of twelve nuclear facilities still remaining (Silos containing uranium ore residues) with its own safety basis documentation. This paper presents the status of the FCP's safety basis documentation program, illustrating that all of the former nuclear facilities and activities have now replaced. Basis of Interim Operations (BIOs) with I-HASPs as their safety basis during the closure process

  20. Review of Policy Documents for Nuclear Safety and Regulation

    International Nuclear Information System (INIS)

    Kim, Woong Sik; Choi, Kwang Sik; Choi, Young Sung; Kim, Hho Jung; Kim, Ho Ki

    2006-01-01

    The goal of regulation is to protect public health and safety as well as environment from radiological hazards that may occur as a result of the use of atomic energy. In September 1994, the Korean government issued the Nuclear Safety Policy Statement (NSPS) to establish policy goals of maintaining and achieving high-level of nuclear safety and also help the public understand the national policy and a strong will of the government toward nuclear safety. It declares the importance of establishing safety culture in nuclear community and also specifies five nuclear regulatory principles (Independence, Openness, Clarity, Efficiency and Reliability) and provides the eleven regulatory policy directions. In 2001, the Nuclear Safety Charter was declared to make the highest goal of safety in driving nuclear business clearer; to encourage atomic energy- related institutions and workers to keep in mind the mission and responsibility for assuring safety; to guarantee public confidence in related organizations. The Ministry of Science and Technology (MOST) also issues Yearly Regulatory Policy Directions at the beginning of every year. Recently, the third Atomic Energy Promotion Plan (2007-2011) has been established. It becomes necessary for the relevant organizations to prepare the detailed plans on such areas as nuclear development, safety management, regulation, etc. This paper introduces a multi-level structure of nuclear safety and regulation policy documents in Korea and presents some improvements necessary for better application of the policies

  1. Review of Policy Documents for Nuclear Safety and Regulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Sik; Choi, Kwang Sik; Choi, Young Sung; Kim, Hho Jung; Kim, Ho Ki [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2006-07-01

    The goal of regulation is to protect public health and safety as well as environment from radiological hazards that may occur as a result of the use of atomic energy. In September 1994, the Korean government issued the Nuclear Safety Policy Statement (NSPS) to establish policy goals of maintaining and achieving high-level of nuclear safety and also help the public understand the national policy and a strong will of the government toward nuclear safety. It declares the importance of establishing safety culture in nuclear community and also specifies five nuclear regulatory principles (Independence, Openness, Clarity, Efficiency and Reliability) and provides the eleven regulatory policy directions. In 2001, the Nuclear Safety Charter was declared to make the highest goal of safety in driving nuclear business clearer; to encourage atomic energy- related institutions and workers to keep in mind the mission and responsibility for assuring safety; to guarantee public confidence in related organizations. The Ministry of Science and Technology (MOST) also issues Yearly Regulatory Policy Directions at the beginning of every year. Recently, the third Atomic Energy Promotion Plan (2007-2011) has been established. It becomes necessary for the relevant organizations to prepare the detailed plans on such areas as nuclear development, safety management, regulation, etc. This paper introduces a multi-level structure of nuclear safety and regulation policy documents in Korea and presents some improvements necessary for better application of the policies.

  2. Preliminary safety assessment and preliminary safety report for the treated radwaste store, Winfrith

    International Nuclear Information System (INIS)

    Staples, A.T.

    1992-06-01

    It is the purpose of this assessment to define the categorisation of the Treated Radwaste Store, TRS, B55 at the Winfrith Technology Centre. Its further purpose is to cover all relevant sections required for a Preliminary Safety Report (PSR) encompassing the TRS and the integral Quality Assessment Unit (QUA). The TRS is designed for the interim storage of intermediate level radioactive wastes. All waste material stored in the TRS will be contained within 500 litre stainless steel drums acceptable to NIREX. It is proposed that the TRS will receive 500 litre stainless steel NIREX drums containing either irradiated DRAGON fuel or encapsulated sludge waste. (author)

  3. CP-50 calibration facility radiological safety assessment document

    International Nuclear Information System (INIS)

    Chilton, M.W.; Hill, R.L.; Eubank, B.F.

    1980-03-01

    The CP-50 Calibration Facility Radiological Safety Assessment document, prepared at the request of the Nevada Operations Office of the US Department of Energy to satisfy provisions of ERDA Manual Chapter 0531, presents design features, systems controls, and procedures used in the operation of the calibration facility. Site and facility characteristics and routine and non-routine operations, including hypothetical incidents or accidents are discussed and design factors, source control systems, and radiation monitoring considerations are described

  4. Hanford Site Wide Transportation Safety Document [SEC 1 Thru 3

    Energy Technology Data Exchange (ETDEWEB)

    MCCALL, D L

    2002-06-01

    This safety evaluation report (SER) documents the basis for the US Department of Energy (DOE), Richland Operations Office (RL) to approve the Hanford Sitewide Transportation Safety Document (TSD) for onsite Transportation and Packaging (T&P) at Hanford. Hanford contractors, on behalf of DOE-RL, prepared and submitted the Hanford Sitewide Transportation Safety Document, DOE/RL-2001-0036, Revision 0, (DOE/RL 2001), dated October 4, 2001, which is referred to throughout this report as the TSD. In the context of the TSD, Hanford onsite shipments are the activities of moving hazardous materials, substances, and wastes between DOE facilities and over roadways where public access is controlled or restricted and includes intra-area and inter-area movements. The TSD sets forth requirements and standards for onsite shipment of radioactive and hazardous materials and wastes within the confines of the Hanford Site on roadways where public access is restricted by signs, barricades, fences, or other means including road closures and moving convoys controlled by Hanford Site security forces.

  5. Preliminary study on improving safety culture in Malaysian nuclear industries

    International Nuclear Information System (INIS)

    Ibrahim, Sabariah Kader; Lee, Y. E.

    2012-01-01

    This paper presents preliminary study on safety culture and its implementation in Malaysian nuclear industries by realizing the importance of safety culture; identification of important safety culture attributes; safety culture assessment and the practices to incorporate the identified safety culture attributes in organization. The first section of this paper explains the terms and definitions related to safety culture. Second, for the realization of importance of safety culture in organization, the international operational experiences emphasizing the importance of safety culture are described. Third, important safety culture attributes which are frequently cited in literature are provided. Fourth, methods to assess safety culture in operating organization are described. Finally, the practices to enhance the safety culture in an organization are discussed

  6. Preliminary study on improving safety culture in Malaysian nuclear industries

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, Sabariah Kader [KAIST, Daejeon (Korea, Republic of); Lee, Y. E. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    This paper presents preliminary study on safety culture and its implementation in Malaysian nuclear industries by realizing the importance of safety culture; identification of important safety culture attributes; safety culture assessment and the practices to incorporate the identified safety culture attributes in organization. The first section of this paper explains the terms and definitions related to safety culture. Second, for the realization of importance of safety culture in organization, the international operational experiences emphasizing the importance of safety culture are described. Third, important safety culture attributes which are frequently cited in literature are provided. Fourth, methods to assess safety culture in operating organization are described. Finally, the practices to enhance the safety culture in an organization are discussed.

  7. Preliminary Safety Analysis Report for the Transuranic Storage Area Retrieval Enclosure at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    1993-03-01

    This Transuranic Storage Area Retrieval Enclosure Preliminary Safety Analysis Report was completed as required by DOE Order 5480.23. The purpose of this document is to construct a safety basis that supports the design and permits construction of the facility. The facility has been designed to the requirements of a Radioactive Solid Waste Facility presented in DOE Order 6430.1A

  8. A preliminary report on OCR problems in LSS document conversion

    International Nuclear Information System (INIS)

    Nartker, T.A.; Kanal, J.; Rice, S.V.

    1992-01-01

    In this paper, the authors present the results of a study to identify the predominant problems which contemporary OCR devices have in converting LSS document images into computer readable text. The results indicate that the majority of residual OCR errors are caused by either broken or touching characters. The relationship between the cost of constructing the LSS system and the number of OCR errors made during LSS data capture is analyzed. if 100% of the conversion problems caused by broken characters alone could be eliminated, the potential LSS cost savings is approximately $31,000,000

  9. Preliminary Design Requirements Document for Project W-314

    Energy Technology Data Exchange (ETDEWEB)

    MCGREW, D.L.

    2000-04-27

    This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1.

  10. Preliminary Design Requirements Document for Project W-314

    International Nuclear Information System (INIS)

    MCGREW, D.L.

    2000-01-01

    This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1

  11. Preliminary Integrated Safety Analysis Status Report

    International Nuclear Information System (INIS)

    Gwyn, D.

    2001-01-01

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

  12. Intranet-based safety documentation in management of major hazards and occupational health and safety.

    Science.gov (United States)

    Leino, Antti

    2002-01-01

    In the European Union, Council Directive 96/82/EC requires operators producing, using, or handling significant amounts of dangerous substances to improve their safety management systems in order to better manage the major accident potentials deriving from human error. A new safety management system for the Viikinmäki wastewater treatment plant in Helsinki, Finland, was implemented in this study. The system was designed to comply with both the new safety liabilities and the requirements of OHSAS 18001 (British Standards Institute, 1999). During the implementation phase experiences were gathered from the development processes in this small organisation. The complete documentation was placed in the intranet of the plant. Hyperlinks between documents were created to ensure convenience of use. Documentation was made accessible for all workers from every workstation.

  13. Planning Document for an NBSR Conversion Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    Diamond D. J.; Baek J.; Hanson, A.L.; Cheng, L-Y.; Brown, N.; Cuadra, A.

    2013-09-25

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted.This report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.

  14. Preliminary Study on the Revision of Nuclear Safety Policy Statement

    International Nuclear Information System (INIS)

    Lee, Y. E.; Lee, S. H.; Chang, H. S.; Choi, K. S.; Jung, S. J.

    2011-01-01

    Nuclear safety policy in Korea is currently declared in the Nuclear Safety Charter as the highest tier document and safety principles and directions are announced in the Nuclear Safety Policy Statement. As the circumstances affecting on the nuclear safety policy change, it needs to revise the Statement. This study aims to develop the revised Nuclear Safety Policy Statement to declare that securing safety is a prerequisite to the utilization of nuclear energy, and that all workers in nuclear industry and regulatory body must adhere to the principle of priority to safety. As a result, two different types of revision are being prepared as of August. One is based on the spirit of Nuclear Safety Charter as well as the direction of future-oriented safety policies including the changes in the environment after declaration of the Statement. The other is to declare the fundamental safety objective and safety principles as the top philosophy of national nuclear safety policy by adopting the '10 Safety Principles in IAEA Safety Fundamental' instead of the current Charter. Both versions of revision are subject to further in-depth discussion. However once the revision is finalized and declared, it would be useful to accomplish effectively the organizational responsibilities and to enhance the public confidence in nuclear safety by performing the regulatory activities in a planned and systematic manner and promulgating the government's dedication to priority to safety

  15. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.

    1995-04-01

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

  16. [Efficiency of computer-based documentation in long-term care--preliminary project].

    Science.gov (United States)

    Lüngen, Markus; Gerber, Andreas; Rupprecht, Christoph; Lauterbach, Karl W

    2008-06-01

    In Germany the documentation of processes in long-term care is mainly paper-based. Planning, realization and evaluation are not supported in an optimal way. In a preliminary study we evaluated the consequences of the introduction of a computer-based documentation system using handheld devices. We interviewed 16 persons before and after introducing the computer-based documentation and assessed costs for the documentation process and administration. The results show that reducing costs is likely. The job satisfaction of the personnel increased, more time could be spent for caring for the residents. We suggest further research to reach conclusive results.

  17. Documented Safety Analysis for the B695 Segment

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-09-11

    This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, or {sup 3}H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building

  18. Documented Safety Analysis for the B695 Segment

    International Nuclear Information System (INIS)

    Laycak, D.

    2008-01-01

    This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., 90 Sr, 137 Cs, or 3 H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building systems, and keeping

  19. Area 5 Radioactive Waste Management Site Safety Assessment Document

    International Nuclear Information System (INIS)

    Horton, K.K.; Kendall, E.W.; Brown, J.J.

    1980-02-01

    The Area 5 Radioactive Waste Management Safety Assessment Document evaluates site characteristics, facilities and operating practices which contribute to the safe handling and storage/disposal of radioactive wastes at the Nevada Test Site. Physical geography, cultural factors, climate and meteorology, geology, hydrology (with emphasis on radionuclide migration), ecology, natural phenomena, and natural resources are discussed and determined to be suitable for effective containment of radionuclides. Also considered, as a separate section, are facilities and operating practices such as monitoring; storage/disposal criteria; site maintenance, equipment, and support; transportation and waste handling; and others which are adequate for the safe handling and storage/disposal of radioactive wastes. In conclusion, the Area 5 Radioactive Waste Management Site is suitable for radioactive waste handling and storage/disposal for a maximum of twenty more years at the present rate of utilization

  20. Documented Safety Analysis for the Waste Storage Facilities March 2010

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  1. Documented Safety Analysis for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  2. Development of Onsite Transportation Safety Documents for Nevada Test Site

    International Nuclear Information System (INIS)

    Frank Hand; Willard Thomas; Frank Sciacca; Manny Negrete; Susan Kelley

    2008-01-01

    Department of Energy (DOE) Orders require each DOE site to develop onsite transportation safety documents (OTSDs). The Nevada Test Site approach divided all onsite transfers into two groups with each group covered by a standalone OTSD identified as Non-Nuclear and Nuclear. The Non-Nuclear transfers involve all radioactive hazardous material in less than Hazard Category (HC)-3 quantities and all chemically hazardous materials. The Nuclear transfers involve all radioactive material equal to or greater than HC-3 quantities and radioactive material mated with high explosives regardless of quantity. Both OTSDs comply with DOE O 460.1B requirements. The Nuclear OTSD also complies with DOE O 461.1A requirements and includes a DOE-STD-3009 approach to hazard analysis (HA) and accident analysis as needed. All Nuclear OTSD proposed transfers were determined to be non-equivalent and a methodology was developed to determine if 'equivalent safety' to a fully compliant Department of Transportation (DOT) transfer was achieved. For each HA scenario, three hypothetical transfers were evaluated: a DOT-compliant, uncontrolled, and controlled transfer. Equivalent safety is demonstrated when the risk level for each controlled transfer is equal to or less than the corresponding DOT-compliant transfer risk level. In this comparison the typical DOE-STD-3009 risk matrix was modified to reflect transportation requirements. Design basis conditions (DBCs) were developed for each non-equivalent transfer. Initial DBCs were based solely upon the amount of material present. Route-, transfer-, and site-specific conditions were evaluated and the initial DBCs revised as needed. Final DBCs were evaluated for each transfer's packaging and its contents

  3. The practical implementation of integrated safety management for nuclear safety analysis and fire hazards analysis documentation

    International Nuclear Information System (INIS)

    COLLOPY, M.T.

    1999-01-01

    the integrated safety management system approach for having a uniform and consistent process: a method has been suggested by the U S . Department of Energy at Richland and the Project Hanford Procedures when fire hazard analyses and safety analyses are required. This process provides for a common basis approach in the development of the fire hazard analysis and the safety analysis. This process permits the preparers of both documents to jointly participate in the development of the hazard analysis process. This paper presents this method to implement the integrated safety management approach in the development of the fire hazard analysis and safety analysis that provides consistency of assumptions. consequences, design considerations, and other controls necessarily to protect workers, the public. and the environment

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

    International Nuclear Information System (INIS)

    SEMMENS, L.S.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    SEMMENS, L.S.

    1999-02-24

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

  6. Classification of Aeronautics System Health and Safety Documents

    Data.gov (United States)

    National Aeronautics and Space Administration — Most complex aerospace systems have many text reports on safety, maintenance, and associated issues. The Aviation Safety Reporting System (ASRS) spans several...

  7. Preliminary safety evaluation for 241-C-106 waste retrieval, project W-320

    International Nuclear Information System (INIS)

    Conner, J.C.

    1994-01-01

    This document presents the Preliminary Safety Evaluation for Project W-320, Tank 241-C-106 Waste Retrieval Sluicing System (WRSS). The US DOE has been mandated to develop plans for response to safety issues associated with the waste storage tanks at the Hanford Site, and to report the progress of implementing those plans to Congress. The objectives of Project W-230 are to design, fabricate, develop, test, and operate a new retrieval system capable of removing a minimum of about 75% of the high-heat waste contained in C-106. It is anticipated that sluicing operations can remove enough waste to reduce the remaining radiogenic heat load to levels low enough to resolve the high-heat safety issue as well as allow closure of the tank safety issue

  8. Preliminary safety analysis report for the Waste Characterization Facility

    International Nuclear Information System (INIS)

    1994-10-01

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

  9. Review of SFR Design Safety using Preliminary Regulatory PSA Model

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2013-01-01

    The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

  10. Hazardous Waste/Mixed Waste Treatment Building Safety Information Document (SID)

    International Nuclear Information System (INIS)

    Fatell, L.B.; Woolsey, G.B.

    1993-01-01

    This Safety Information Document (SID) provides a description and analysis of operations for the Hazardous Waste/Mixed Waste Disposal Facility Treatment Building (the Treatment Building). The Treatment Building has been classified as a moderate hazard facility, and the level of analysis performed and the methodology used are based on that classification. Preliminary design of the Treatment Building has identified the need for two separate buildings for waste treatment processes. The term Treatment Building applies to all these facilities. The evaluation of safety for the Treatment Building is accomplished in part by the identification of hazards associated with the facility and the analysis of the facility's response to postulated events involving those hazards. The events are analyzed in terms of the facility features that minimize the causes of such events, the quantitative determination of the consequences, and the ability of the facility to cope with each event should it occur. The SID presents the methodology, assumptions, and results of the systematic evaluation of hazards associated with operation of the Treatment Building. The SID also addresses the spectrum of postulated credible events, involving those hazards, that could occur. Facility features important to safety are identified and discussed in the SID. The SID identifies hazards and reports the analysis of the spectrum of credible postulated events that can result in the following consequences: Personnel exposure to radiation; Radioactive material release to the environment; Personnel exposure to hazardous chemicals; Hazardous chemical release to the environment; Events leading to an onsite/offsite fatality; and Significant damage to government property. The SID addresses the consequences to the onsite and offsite populations resulting from postulated credible events and the safety features in place to control and mitigate the consequences

  11. Hazardous Waste/Mixed Waste Treatment Building Safety Information Document (SID)

    Energy Technology Data Exchange (ETDEWEB)

    Fatell, L.B.; Woolsey, G.B.

    1993-04-15

    This Safety Information Document (SID) provides a description and analysis of operations for the Hazardous Waste/Mixed Waste Disposal Facility Treatment Building (the Treatment Building). The Treatment Building has been classified as a moderate hazard facility, and the level of analysis performed and the methodology used are based on that classification. Preliminary design of the Treatment Building has identified the need for two separate buildings for waste treatment processes. The term Treatment Building applies to all these facilities. The evaluation of safety for the Treatment Building is accomplished in part by the identification of hazards associated with the facility and the analysis of the facility`s response to postulated events involving those hazards. The events are analyzed in terms of the facility features that minimize the causes of such events, the quantitative determination of the consequences, and the ability of the facility to cope with each event should it occur. The SID presents the methodology, assumptions, and results of the systematic evaluation of hazards associated with operation of the Treatment Building. The SID also addresses the spectrum of postulated credible events, involving those hazards, that could occur. Facility features important to safety are identified and discussed in the SID. The SID identifies hazards and reports the analysis of the spectrum of credible postulated events that can result in the following consequences: Personnel exposure to radiation; Radioactive material release to the environment; Personnel exposure to hazardous chemicals; Hazardous chemical release to the environment; Events leading to an onsite/offsite fatality; and Significant damage to government property. The SID addresses the consequences to the onsite and offsite populations resulting from postulated credible events and the safety features in place to control and mitigate the consequences.

  12. IAEA activities in preparation of reglamentary documents on nuclear power plant safety

    International Nuclear Information System (INIS)

    Konstantinov, L.V.

    1976-01-01

    The activities of the IAEA in the field of working out practical rules and recommendations ensuring the nuclear power plant safety are discussed. The practical rules will establish the aims and the minimum of requirements, that must be carried out to ensure the necessary safety of systems, components and equipment of the nuclear power plant throughout the whole period of its exploitation. Described is the procedure of the document preparation, consisting of the collection of documents, edited in different countries, the integration of documents by the IAEA Secretariat, the consideratiom of documents by the Group of senior advisers, the preparation of the draft document, the additional wort at the document in accordaqce with the remarks of the IAEA member-countries, the edition and dissemination of documents. The necessity for the active participation of the CMEA member-countries in the development and discussion of documents concerning the nuclear power plant safety is stated [ru

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

    International Nuclear Information System (INIS)

    Batchelor, K.

    1996-03-01

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

  14. 33 CFR 96.250 - What documents and reports must a safety management system have?

    Science.gov (United States)

    2010-07-01

    ... safety management system have? 96.250 Section 96.250 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.250 What documents and...

  15. Using resources for scientific-driven pharmacovigilance: from many product safety documents to one product safety master file.

    Science.gov (United States)

    Furlan, Giovanni

    2012-08-01

    Current regulations require a description of the overall safety profile or the specific risks of a drug in multiple documents such as the Periodic and Development Safety Update Reports, Risk Management Plans (RMPs) and Signal Detection Reports. In a resource-constrained world, the need for preparing multiple documents reporting the same information results in shifting the focus from a thorough scientific and medical evaluation of the available data to maintaining compliance with regulatory timelines. Since the aim of drug safety is to understand and characterize product issues to take adequate risk minimization measures rather than to comply with bureaucratic requirements, there is the need to avoid redundancy. In order to identify core drug safety activities that need to be undertaken to protect patient safety and reduce the number of documents reporting the results of these activities, the author has reviewed the main topics included in the drug safety guidelines and templates. The topics and sources that need to be taken into account in the main regulatory documents have been found to greatly overlap and, in the future, as a result of the new Periodic Safety Update Report structure and requirements, in the author's opinion this overlap is likely to further increase. Many of the identified inter-document differences seemed to be substantially formal. The Development Safety Update Report, for example, requires separate presentation of the safety issues emerging from different sources followed by an overall evaluation of each safety issue. The RMP, instead, requires a detailed description of the safety issues without separate presentation of the evidence derived from each source. To some extent, however, the individual documents require an in-depth analysis of different aspects; the RMP, for example, requires an epidemiological description of the indication for which the drug is used and its risks. At the time of writing this article, this is not specifically

  16. Consensus document regarding cardiovascular safety at sports arenas

    DEFF Research Database (Denmark)

    Borjesson, Mats; Serratosa, Luis; Carre, Francois

    2011-01-01

    Mass gathering events in sports arenas create challenges regarding the cardiovascular safety of both athletes and spectators. A comprehensive medical action plan, to ensure properly applied cardiopulmonary resuscitation, and wide availability and use of automated external defibrillators (AEDs......), is essential to improving survival from sudden cardiac arrest at sporting events. This paper outlines minimum standards for cardiovascular care to assist in the planning of mass gathering sports events across Europe with the intention of local adaptation at individual sports arenas, to ensure the full...

  17. Safety assessment document for the dynamic test complex (Building 836)

    International Nuclear Information System (INIS)

    Odell, B.N.; Pfeifer, H.E.

    1981-01-01

    A safety assessment was performed to determine if potential accidents at the 836 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (H) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire

  18. Safety assessment document for the dynamic test complex (Building 836)

    Energy Technology Data Exchange (ETDEWEB)

    Odell, B.N.; Pfeifer, H.E.

    1981-11-24

    A safety assessment was performed to determine if potential accidents at the 836 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (H) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire.

  19. Approaches to document the efficacy and safety of microdermabrasion procedure.

    Science.gov (United States)

    Spencer, James M; Kurtz, Ellen S

    2006-11-01

    Microdermabrasion is a popular cosmetic procedure for skin rejuvenation, which is achieved by mechanical abrasion of the skin at a superficial level. The objective was to study the efficacy of microdermabrasion in photoaging and to investigate the compatibility of a cleanser and a lotion with microdermabrasion. Sixteen women underwent microdermabrasion to the face once a week for a total of six treatments. Subjects were also given a personal skin care regimen (cleanser and lotion). Colorimetry values as well as investigator and patients ratings for safety and efficacy were analyzed at various time points and compared to baseline. There were no significant changes in safety and tolerance variables throughout the study. Clinical efficacy variables (fine wrinkles, dullness, pigmentation, and large pores) significantly improved by the third treatment, with further improvement by the end of the study (six treatments). Subjects perceived improvement in facial photoaging variables. Colorimetry showed increased brightness and decreased yellowness of target skin sites on the face throughout the study. Multiple microdermabrasions were effective in significantly improving various facial photoaging variables. The personal skin care regimen used was well tolerated by the subjects.

  20. Safety assessment document for the Dynamic Test Complex B854

    International Nuclear Information System (INIS)

    Odell, B.N.; Pfeifer, H.E.

    1981-01-01

    A safety assessment was performed to determine if potential accidents at the 854 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (HE) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire. Seismic and extreme wind (including missiles) analyses indicate that the buildings are basically sound. The lightning protection system is in the process of being upgraded to meet AMCR 385-100. These buildings are located high above the dry creek bed so that a flood is improbable. The probability of high explosive detonation involving plutonium is very remote since the radioactive materials are encased and plutonium and HE are not permitted concurrently in the same area at Site 300. (The exception to this policy is that explosive actuating devices are sometimes located in assemblies containing fissile materials. However, an accidental actuation will not affect the safe containment of the plutonium within the assembly.) There is a remote possibility of an HE explosion involving uranium and beryllium since these are permitted in the same area.The possibility of a criticality accident is very remote since the fissile materials are doubly encased in stout metal containers. All operations involving these materials are independently reviewed and inspected by the Criticality Safety Office. It was determined that a fire was unlikely due to the low fire loading and the absence of ignition sources. It was also determined that the consequences of any accidents were reduced by the remote location of these facilities, their design, and by administrative controls

  1. Preliminary safety evaluation for the plutonium stabilization and packaging system

    International Nuclear Information System (INIS)

    Shapley, J.E.

    1997-01-01

    This Preliminary Safety Evaluation (PSE) describes and analyzes the installation and operation of the Plutonium Stabilization and Packaging System (SPS) at the Plutonium Finishing Plant (PFP). The SPS is a combination of components required to expedite the safe and timely storage of Plutonium (Pu) oxide. The SPS program will receive site Pu packages, process the Pu for storage, package the Pu into metallic containers, and safely store the containers in a specially modified storage vault. The location of the SPS will be in the 2736- ZB building and the storage vaults will be in the 2736-Z building of the PFP, as shown in Figure 1-1. The SPS will produce storage canisters that are larger than those currently used for Pu storage at the PFP. Therefore, the existing storage areas within the PFP secure vaults will require modification. Other modifications will be performed on the 2736-ZB building complex to facilitate the installation and operation of the SPS

  2. BANDUNG JENGKI FROM HERITAGE POINT OF VIEW: DOCUMENTATION AND PRELIMINARY SEARCH ON SIGNIFICANCE

    Directory of Open Access Journals (Sweden)

    WULANDARI Ratri

    2016-07-01

    Full Text Available Jengki architecture is one of Indonesian post-colonial architecture style popular in 1950 to 1970. Most study of Jengki was done in Jakarta, Semarang, and Malang. In Bandung, there is no written record of Jengki identification or documentation found. This paper tries to recognize and document Jengki typology in Bandung as a preliminary study on its presences, and its variant, in the city. It also tries to look at this typology from heritage point of view, whether there is significance in accordance to heritage criteria. Using site survey and Google street view, data were collected and elaborated in table to recognize Jengki traits in a suspected building. Building’s significances were then reviewed using criteria mentioned in local legislation.  It is found that not all suspected Jengki building contain all Jengki traits rather more late modern Indonesian character.  Private buildings were assessed to have less heritage value, while public building retains more. Deeper study and documentation on Jengki in Bandung are still needed to legitimate Jengki as heritage since there is no study ever done previously on this issue

  3. ESRS guidelines for software safety reviews. Reference document for the organization and conduct of Engineering Safety Review Services (ESRS) on software important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    The IAEA provides safety review services to assist Member States in the application of safety standards and, in particular, to evaluate and facilitate improvements in nuclear power plant safety performance. Complementary to the Operational Safety Review Team (OSART) and the International Regulatory Review Team (IRRT) services are the Engineering Safety Review Services (ESRS), which include reviews of siting, external events and structural safety, design safety, fire safety, ageing management and software safety. Software is of increasing importance to safety in nuclear power plants as the use of computer based equipment and systems, controlled by software, is increasing in new and older plants. Computer based devices are used in both safety related applications (such as process control and monitoring) and safety critical applications (such as reactor protection). Their dependability can only be ensured if a systematic, fully documented and reviewable engineering process is used. The ESRS on software safety are designed to assist a nuclear power plant or a regulatory body of a Member State in the review of documentation relating to the development, application and safety assessment of software embedded in computer based systems important to safety in nuclear power plants. The software safety reviews can be tailored to the specific needs of the requesting organization. Examples of such reviews are: project planning reviews, reviews of specific issues and reviews prior final acceptance. This report gives information on the possible scope of ESRS software safety reviews and guidance on the organization and conduct of the reviews. It is aimed at Member States considering these reviews and IAEA staff and external experts performing the reviews. The ESRS software safety reviews evaluate the degree to which software documents show that the development process and the final product conform to international standards, guidelines and current practices. Recommendations are

  4. Implementing 10 CFR 830 at the FEMP Silos: Nuclear Health and Safety Plans as Documented Safety Analysis

    International Nuclear Information System (INIS)

    Fisk, Patricia; Rutherford, Lavon

    2003-01-01

    The objective of the Silos Project at the Fernald Closure Project (FCP) is to safely remediate high-grade uranium ore residues (Silos 1 and 2) and metal oxide residues (Silo 3). The evolution of Documented Safety Analyses (DSAs) for these facilities has reflected the changes in remediation processes. The final stage in silos DSAs is an interpretation of 10 CFR 830 Safe Harbor Requirements that combines a Health and Safety Plan with nuclear safety requirements. This paper will address the development of a Nuclear Health and Safety Plan, or N-HASP

  5. Implementing national nuclear safety plan at the preliminary stage of nuclear power project development

    International Nuclear Information System (INIS)

    Xue Yabin; Cui Shaozhang; Pan Fengguo; Zhang Lizhen; Shi Yonggang

    2014-01-01

    This study discusses the importance of nuclear power project design and engineering methods at the preliminary stage of its development on nuclear power plant's operational safety from the professional view. Specifically, we share our understanding of national nuclear safety plan's requirement on new reactor accident probability, technology, site selection, as well as building and improving nuclear safety culture and strengthening public participation, with a focus on plan's implications on preliminary stage of nuclear power project development. Last, we introduce China Huaneng Group's work on nuclear power project preliminary development and the experience accumulated during the process. By analyzing the siting philosophy of nuclear power plant and the necessity of building nuclear safety culture at the preliminary stage of nuclear power project development, this study explicates how to fully implement the nuclear safety plan's requirements at the preliminary stage of nuclear power project development. (authors)

  6. Preliminary safety analysis of molten salt breeder reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2013-01-01

    Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

  7. Keeping rail on track: preliminary findings on safety culture in Australian rail.

    Science.gov (United States)

    Blewett, Verna; Rainbird, Sophia; Dorrian, Jill; Paterson, Jessica; Cattani, Marcus

    2012-01-01

    'Safety culture' is identified in the literature as a critical element of healthy and safe workplaces. How can rail organizations ensure that consistently effective work health and safety cultures are maintained across the diversity of their operations? This paper reports on research that is currently underway in the Australian rail industry aimed at producing a Model of Best Practice in Safety Culture for the industry. Located in rail organizations dedicated to the mining industry as well as urban rail and national freight operations, the research examines the constructs of organizational culture that impact on the development and maintenance of healthy and safe workplaces. The research uses a multi-method approach incorporating quantitative (survey) and qualitative (focus groups, interviews and document analysis) methods along with a participative process to identify interventions to improve the organization and develop plans for their implementation. The research uses as its analytical framework the 10 Platinum Rules, from the findings of earlier research in the New South Wales (Australia) mining industry, Digging Deeper. Data collection is underway at the time of writing and preliminary findings are presented at this stage. The research method may be adapted for use as a form of organizational review of safety and health in organizational culture.

  8. Health Information Technology, Patient Safety, and Professional Nursing Care Documentation in Acute Care Settings.

    Science.gov (United States)

    Lavin, Mary Ann; Harper, Ellen; Barr, Nancy

    2015-04-14

    The electronic health record (EHR) is a documentation tool that yields data useful in enhancing patient safety, evaluating care quality, maximizing efficiency, and measuring staffing needs. Although nurses applaud the EHR, they also indicate dissatisfaction with its design and cumbersome electronic processes. This article describes the views of nurses shared by members of the Nursing Practice Committee of the Missouri Nurses Association; it encourages nurses to share their EHR concerns with Information Technology (IT) staff and vendors and to take their place at the table when nursing-related IT decisions are made. In this article, we describe the experiential-reflective reasoning and action model used to understand staff nurses' perspectives, share committee reflections and recommendations for improving both documentation and documentation technology, and conclude by encouraging nurses to develop their documentation and informatics skills. Nursing issues include medication safety, documentation and standards of practice, and EHR efficiency. IT concerns include interoperability, vendors, innovation, nursing voice, education, and collaboration.

  9. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  10. Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

    International Nuclear Information System (INIS)

    Lydell, B.

    1998-03-01

    This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood

  11. Preliminary safety evaluation for a medical therapy reactor

    International Nuclear Information System (INIS)

    Jones, J.L.; Neuman, W.A.

    1989-01-01

    A conceptual design of a passively safe reactor facility for boron neutron capture therapy has been previously described. The medical therapy reactor (MTR) has a maximum power level of 10 MW(thermal) and utilizes 45 wt% uranium in UZrH, 20 wt% 235 U enriched hydride fuel matrix with 1 wt% erbium, which is a burnable poison and provides prompt negative reactivity feedback. The facility has five beam ports for patient treatment and advanced neutron beam research and is capable of 2,000 to 10,000 treatments per year, assuming single 8h/day, 5 day/week operation. The epithermal treatment flux from the beam ports is large, enabling single-session treatment of brain cancers of <10-min duration, with minimal fast neutron and gamma contaminants. The reactor core is designed with sufficient excess reactivity to yield a core lifetime equal to a facility lifetime of 30 yr. A preliminary safety evaluation was performed using the RELAP5 thermal-hydraulic code. The analysis addressed accidents in several major categories, including a pump coastdown, a loss of secondary heat sink, and a $0.5 step reactivity insertion

  12. Preliminary safety evaluation for the spent nuclear fuel project`s cold vacuum drying system

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J., Westinghouse Hanford

    1996-07-01

    This preliminary safety evaluation (PSE) considers only the Cold Vacuum Drying System (CVDS) facility and its mission as it relates to the integrated process strategy (WHC 1995). The purpose of the PSE is to identify those CBDS design functions that may require safety- class and safety-significant accident prevention and mitigation features.

  13. Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

    International Nuclear Information System (INIS)

    Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

    1990-01-01

    The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

  14. Electronic Medical Record Documentation of Driving Safety for Veterans with Diagnosed Dementia.

    Science.gov (United States)

    Vair, Christina L; King, Paul R; Gass, Julie; Eaker, April; Kusche, Anna; Wray, Laura O

    2018-01-01

    Many older adults continue to drive following dementia diagnosis, with medical providers increasingly likely to be involved in addressing such safety concerns. This study examined electronic medical record (EMR) documentation of driving safety for veterans with dementia (N = 118) seen in Veterans Affairs primary care and interdisciplinary geriatrics clinics in one geographic region over a 10-year period. Qualitative directed content analysis of retrospective EMR data. Assessment of known risk factors or subjective concerns for unsafe driving were documented in fewer than half of observed cases; specific recommendations for driving safety were evident for a minority of patients, with formal driving evaluation the most frequently documented recommendation by providers. Utilizing data from actual clinical encounters provides a unique snapshot of how driving risk and safety concerns are addressed for veterans with dementia. This information provides a meaningful frame of reference for understanding potential strengths and possible gaps in how this important topic area is being addressed in the course of clinical care. The EMR is an important forum for interprofessional communication, with documentation of driving risk and safety concerns an essential element for continuity of care and ensuring consistency of information delivered to patients and caregivers.

  15. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.

  16. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    International Nuclear Information System (INIS)

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included

  17. Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-10-01

    Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.

  18. IMPLEMENTING CHANGES TO AN APPROVED AND IN-USE DOCUMENTED SAFETY ANALYSIS

    International Nuclear Information System (INIS)

    KING JP

    2008-01-01

    The Plutonium Finishing Plant (PFP) has refined a process to ensure a comprehensive and complete DSA/TSR change implementation. Successful Nuclear Facility Safety Basis implementation is essential to avoid creating a Potential Inadequacy in Safety Analysis (PISA) situation, or implementing a facility into a non-compliance that can result in a TSR violation. Once past initial implementation, additional changes to Documented Safety Analysis (DSA) and Technical Safety Requirements (TSRs) are often needed due to needed requirement clarifications, operating experience indicating that Conditions/Required Actions/Surveillance Requirements could be improved, changes in facility conditions, or changes in facility mission etc. An effective change implementation process is essential to ensuring compliance with 10 CFR 830.202(a), 'The contractor responsible for a hazard category 1,2, or 3 DOE nuclear facility must establish and maintain the safety basis for the facility'

  19. Preliminary report of radiological safety to hydrology 1993 campaign

    International Nuclear Information System (INIS)

    Badano, A.; Suarez Antola, R.; Dellepere, A.; Barreiro, M.

    1993-01-01

    This report has been prepared based on the interaction between project managers and division radiological Protection and Nuclear Safety. In seeking to establish a basis for approval from the point of view of radiation safety practices . The idea for the audit has been provided at all times because the interest was the exchange of ideas and the use of common sense to improve the safety of radioactive substances, security of operators and public safety and environment.The above shows that in the planned radiation safety condition described in this report,the practice can be carried out according to the criteria of safety accepted .

  20. 78 FR 13563 - Energy Conservation Program: Availability of the Preliminary Technical Support Document for...

    Science.gov (United States)

    2013-02-28

    ... identify and resolve issues involved in the preliminary analyses. Chapter 2 of the preliminary technical... DOE conducted in-depth technical analyses in the following areas for GSFLs and IRLs currently under... also begun work on the manufacturer impact analysis and identified the methods to be used for the LCC...

  1. Synthesis of the IRSN report on its analysis of the safety guidance package (DOrS) of the ASTRID reactor project. Safety guidance document for the ASTRID prototype: Referral to the GPR. Opinion related to the safety guidance document of the ASTRID reactor project. ASTRID prototype: Safety guidance document for the ASTRID prototype

    International Nuclear Information System (INIS)

    Lachaume, Jean-Luc; Niel, Jean-Christophe

    2013-01-01

    A first document indicates the improvement guidelines for the ASTRID project based on the French experience in the field of sodium-cooled fast neutron reactors, addresses the safety objectives as they are presented for the ASTRID project, discusses how the project includes a regulation and design referential, and how it addresses various aspects of the design approach (ranking and analysis of operation situations, defence in depth, use of probabilistic studies, safety classification and qualification to accidental situations, taking internal and external aggressions into account and taking severe accidents into account at the design level). It comments the guidelines related to the first two barriers, to main safety functions (control of reactivity and of reactor cooling, containment of radioactive and toxic materials), to dismantling, to R and D for safety support. A second document is a letter sent by the ASN to the GPR (permanent group of experts in charge of nuclear reactors) about the safety guidance document for the ASTRID prototype. The third document is the answer and contains comments and recommendations by this group about the content of this document, and therefore addresses the same topics as the first document. The last document defines the framework of the approach to this document

  2. ITER final design report, cost review and safety analysis (FDR) and relevant documents

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains the fourth major milestone report and documents associated with its acceptance, review and approval. This ITER Final Design Report, Cost Review and Safety Analysis was presented to the ITER Council at its 13th meeting in February 1998 and was approved at its extraordinary meeting on 25 June 1998. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics and schedule and cost estimates are given

  3. Environmental qualification - walkdowns: The documentation of configuration information for safety related components, equipment and systems

    International Nuclear Information System (INIS)

    Melmer, J.; Waters, M.

    1995-01-01

    Environmental Qualification walkdowns are conducted to collect field data to verify/validate/document configurations of safety related equipment and systems. This paper describes the process for conducting walkdowns and the justification for using an electronic format. The following are described: a) Background; b) Preparing, executing and processing walkdowns; c) Hardware/software; d) Impact of a paperless system on walkdown execution, maintenance and work planning; e) Other applications for the technology

  4. THE FORMATION OF THE CONTOUR OF THE DOCUMENTED AND REAL FLIGHT SAFETY IN THE SYSTEM OF THE INFORMATION PROVISION OF SAFETY OF FLIGHTS

    Directory of Open Access Journals (Sweden)

    B. I. Bachkalo

    2015-01-01

    Full Text Available The article discusses the principles and mechanisms of formation of the contour of the real safety of flights and contour of the documented safety, allowing us to obtain information to control fligh safety. The proposed approach can be used in the algorithms of active on-board flight safety management system for the implementation of information support to the crew in flight and automatic control of flight safety.

  5. Preliminary report in radiological safety for 1993 hydrology campaign

    International Nuclear Information System (INIS)

    Badano, A.; Suraez, R.; Dellepere, A.; Barreiro, M.

    1993-01-01

    The purpose of this report is to provide a study about industrial effluents influence on water pollution of Montevideo coastal beaches. The methods which have been considered are nuclear tracer techniques with a special attention in the radioprotection supervision. Three points are considered as evaluation: handling of radioactive tracers and safety, radiation protection workers, environment and public safety. tabs

  6. LESSONS LEARNED IN DEVELOPMENT OF THE HANFORD SWOC MASTER DOCUMENTED SAFETY ANALYSIS (MDSA) and IMPLEMENTATION VALIDATION REVIEW (IVR)

    International Nuclear Information System (INIS)

    MORENO, M.R.

    2004-01-01

    DOE set clear expectations on a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (20 CFR 830, Nuclear Safety Rule), which ensured long-term benefit to Hanford, via issuance of a nuclear safety strategy in February 2003. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development with the goal of a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was approved to standardize methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was approved for the evaluation of radiological consequences for accident scenarios often postulated at Hanford. Standard safety management program chapters were approved for use as a means of compliance with the programmatic chapters of DOE-STD-3009, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports''. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. The new Documented Safety Analysis (DSA) developed to address the operations of four facilities within the Solid Waste Operations Complex (SWOC) necessitated development of an Implementation Validation Review (IVR) process. The IVR process encompasses the following objectives: safety basis controls and requirements are adequately incorporated into appropriate facility documents and work instructions, facility personnel are knowledgeable of controls and requirements, and the DSA/TSR controls have been implemented. Based on DOE direction and safety analysis tools, four waste management nuclear facilities were integrated into one safety basis document. With successful completion of implementation of this safety document, lessons-learned from the in-process review, safety analysis tools and IVR process were documented for future action

  7. 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)

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

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1997-01-01

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

  9. Recommendations: Procedure to develop a preliminary safety report as part of the radioactive waste repository construction licensing process

    International Nuclear Information System (INIS)

    2003-01-01

    The structure of a preliminary safety report for the title purpose should be as follows: A. Textual part: 1. General (Introduction, Basic information about the construction, Timetable); 2. Site information (Siting, Geography and demography, Meteorology and climatic situation, Hydrology, Geology and hydrogeology); 3. Repository design description (Basic function and performance requirements, Design, Auxiliary systems, Fire prevention/protection, Emergency plans); 4. Operation of the repository (Waste acceptance and inspection, Waste handling and interim storage, Waste disposal, Operating monitoring), 5. Health and environmental impact assessment (Radionuclide inventory, Radionuclide transport paths and mechanisms of release into the environment, Radionuclide release in normal and emergency situations, Radiation protection - health impact assessment and regulatory compliance, Draft operating limits and conditions, Proposed ways of assuring physical protection, Uncertainty assessment), 6. Safe repository shutdown/decommissioning concept, 7 Quality assurance assessment, 8. List of selected equipment. B. Annexes: Maps, Drawings, Diagrams, Miscellaneous; C. Documentation: Previous safety report amendments, Protocols, Miscellaneous. (P.A.)

  10. Preliminary Study on the Development of Quantitative Safety Culture Index

    International Nuclear Information System (INIS)

    Lee, Young Eal; Kim, Hun Sil; Ahn, Nam Sung

    2005-01-01

    Safety culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. Because it needs to be recognized as the most significant consciousness to achieve the nuclear safety performance, Korean government and nuclear power generation company have tried to develop the practical method to improve the safety culture from the long term point view. In this study, based on the site interviews to define the potential issues on organizational behavior for the safe operation and the survey on the level of safety culture of occupied workers are conducted. Survey results are quantified as a few indicators of nuclear safety by the statistical method and it can be simulated by the dynamic modeling as time goes on. Currently index and dynamic modeling are still being developed, however, results can be used to suggest the long term strategy which safety is clearly integrated into all activities in the nuclear organization

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

    International Nuclear Information System (INIS)

    Salazar, R.J.; Lane, S.

    1992-02-01

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

  12. Preliminary design requirements document (DRD) for Project W-236B, ''Initial Pretreatment Module''

    International Nuclear Information System (INIS)

    Swanson, L.M.

    1995-01-01

    The scope of this Design Requirements Document (DRD) is to identify and define the functions, with associated requirements, which must be performed to separate Hanford Site tank waste supernatants into low-level and high-level fractions. This documents sets forth function requirements, performance requirements, and design constraints necessary to begin conceptual design for the Initial Pretreatment Module (IPM). System and physical interfaces between the IPM project and the Tank Waste Remediation System (TWRS) are identified. The constraints, performance requirements, and transfer of information and data across a technical interface will be documented in an Interface Control Document. Supplemental DRDs will be prepared to provide more detailed requirements specific to systems described in the DRD

  13. 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

  14. A document-driven method for certifying scientific computing software for use in nuclear safety analysis

    International Nuclear Information System (INIS)

    Smith, W. Spencer; Koothoor, Mimitha

    2016-01-01

    This paper presents a documentation and development method to facilitate the certification of scientific computing software used in the safety analysis of nuclear facilities. To study the problems faced during quality assurance and certification activities, a case study was performed on legacy software used for thermal analysis of a fuel pin in a nuclear reactor. Although no errors were uncovered in the code, 27 issues of incompleteness and inconsistency were found with the documentation. This work proposes that software documentation follow a rational process, which includes a software requirements specification following a template that is reusable, maintainable, and understandable. To develop the design and implementation, this paper suggests literate programming as an alternative to traditional structured programming. Literate programming allows for documenting of numerical algorithms and code together in what is termed the literate programmer's manual. This manual is developed with explicit traceability to the software requirements specification. The traceability between the theory, numerical algorithms, and implementation facilitates achieving completeness and consistency, as well as simplifies the process of verification and the associated certification

  15. A document-driven method for certifying scientific computing software for use in nuclear safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Smith, W. Spencer; Koothoor, Mimitha [Computing and Software Department, McMaster University, Hamilton (Canada)

    2016-04-15

    This paper presents a documentation and development method to facilitate the certification of scientific computing software used in the safety analysis of nuclear facilities. To study the problems faced during quality assurance and certification activities, a case study was performed on legacy software used for thermal analysis of a fuel pin in a nuclear reactor. Although no errors were uncovered in the code, 27 issues of incompleteness and inconsistency were found with the documentation. This work proposes that software documentation follow a rational process, which includes a software requirements specification following a template that is reusable, maintainable, and understandable. To develop the design and implementation, this paper suggests literate programming as an alternative to traditional structured programming. Literate programming allows for documenting of numerical algorithms and code together in what is termed the literate programmer's manual. This manual is developed with explicit traceability to the software requirements specification. The traceability between the theory, numerical algorithms, and implementation facilitates achieving completeness and consistency, as well as simplifies the process of verification and the associated certification.

  16. 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)

  17. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  18. Preliminary Marine Safety Risk Assessment, Brandon Road Lock and Dam Invasive Species Control Measures

    Science.gov (United States)

    2016-12-01

    Decision makers must include control-measure monitoring and emergency “interventions” to insure safety. The Coast Guard operational commanders...system” incorporates a travelling car on a rail above the barge-loading wharf to prevent loading personnel, cargo surveyors, or others from falling...to the Gulf of Mexico . As “Loopers”, they will have already transited the CSSC electric barriers. Preliminary Marine Safety Risk Assessment, BRLD

  19. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  20. Analysis of compatibility of current Czech initial documentation in the area of technical assurance of nuclear safety with the requirements of the EUR document

    International Nuclear Information System (INIS)

    Zdebor, J.; Zdebor, R.; Kratochvil, L.

    2001-11-01

    The publication is structured as follows: Description of existing documentation. General requirements, goals, principles and design principles: Documents being compared; Method of comparison; Results and partial evaluation of comparison of requirements between EUR and Czech regulations (basic goals and safety philosophy; quantitative safety objectives; basic design requirements; extended design requirements; external and internal threats; technical requirements; site conditions); Summary of the comparison of safety requirements. Comparison of requirements for the systems: Requirements for the nuclear reactor unit systems; Barrier systems (fuel system; reactor cooling system; containment system); Remaining systems (control systems; protection systems; coolant makeup and purification system; residual heat removal system; emergency cooling system; power systems); Common technical requirements for systems (technical requirements for systems; internal and external events). (P.A.)

  1. Preliminary application of Structure from Motion and GIS to document decomposition and taphonomic processes.

    Science.gov (United States)

    Carlton, Connor D; Mitchell, Samantha; Lewis, Patrick

    2018-01-01

    Over the past decade, Structure from Motion (SfM) has increasingly been used as a means of digital preservation and for documenting archaeological excavations, architecture, and cultural material. However, few studies have tapped the potential of using SfM to document and analyze taphonomic processes affecting burials for forensic sciences purposes. This project utilizes SfM models to elucidate specific post-depositional events that affected a series of three human cadavers deposited at the South East Texas Applied Forensic Science Facility (STAFS). The aim of this research was to test the ability for untrained researchers to employ spatial software and photogrammetry for data collection purposes. For a series of three months a single lens reflex (SLR) camera was used to capture a series of overlapping images at periodic stages in the decomposition process of each cadaver. These images are processed through photogrammetric software that creates a 3D model that can be measured, manipulated, and viewed. This project used photogrammetric and geospatial software to map changes in decomposition and movement of the body from original deposition points. Project results indicate SfM and GIS as a useful tool for documenting decomposition and taphonomic processes. Results indicate photogrammetry is an efficient, relatively simple, and affordable tool for the documentation of decomposition. Copyright © 2017 Elsevier B.V. All rights reserved.

  2. Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1998-03-01

    This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood. 7 refs, 4 tabs, 3 figs. Also available at the SKI Home page: //www.ski.se.

  3. Preliminary considerations on safety of computerized control rooms

    International Nuclear Information System (INIS)

    Vittet, J.

    1983-02-01

    Safety problems are analyzed in this report by the study of the interaction: ''human behavior in a rigid environment/information overload in perturbed situation''. For pedagogy the study is presented as a research of factors influencing operator performance in a control room and a dialogue between an analyst and a conceiving engineer. Danger of all control room where the strategy for data acquisition is too rigid and without spatial reference is stressed in conclusion. Orientations for an advanced control room are outlined [fr

  4. Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Perdomo, Manuel

    1995-01-01

    Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential 'weak points' at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs

  5. Safety of High Speed Magnetic Levitation Transportation Systems: Preliminary Safety Review of the Transrapid Maglev System

    Science.gov (United States)

    1990-11-01

    The safety of various magnetically levitated trains under development for possible : implementation in the United States is of direct concern to the Federal Railroad : Administration. This report, one in a series of planned reports on maglev safety, ...

  6. Preliminary safety evaluation (PSE) for Sodium Storage Facility at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bowman, B.R.

    1994-01-01

    This evaluation was performed for the Sodium Storage Facility (SSF) which will be constructed at the Fast Flux Test Facility (FFTF) in the area adjacent to the South and West Dump Heat Exchanger (DHX) pits. The purpose of the facility is to allow unloading the sodium from the FFTF plant tanks and piping. The significant conclusion of this Preliminary Safety Evaluation (PSE) is that the only Safety Class 2 components are the four sodium storage tanks and their foundations. The building, because of its imminent risk to the tanks under an earthquake or high winds, will be Safety Class 3/2, which means the building has a Safety Class 3 function with the Safety Class 2 loads of seismic and wind factored into the design

  7. Ibogaine: complex pharmacokinetics, concerns for safety, and preliminary efficacy measures.

    Science.gov (United States)

    Mash, D C; Kovera, C A; Pablo, J; Tyndale, R F; Ervin, F D; Williams, I C; Singleton, E G; Mayor, M

    2000-09-01

    Ibogaine is an indole alkaloid found in the roots of Tabernanthe Iboga (Apocynaceae family), a rain forest shrub that is native to western Africa. Ibogaine is used by indigenous peoples in low doses to combat fatigue, hunger and thirst, and in higher doses as a sacrament in religious rituals. Members of American and European addict self-help groups have claimed that ibogaine promotes long-term drug abstinence from addictive substances, including psychostimulants and opiates. Anecdotal reports attest that a single dose of ibogaine eliminates opiate withdrawal symptoms and reduces drug craving for extended periods of time. The purported efficacy of ibogaine for the treatment of drug dependence may be due in part to an active metabolite. The majority of ibogaine biotransformation proceeds via CYP2D6, including the O-demethylation of ibogaine to 12-hydroxyibogamine (noribogaine). Blood concentration-time effect profiles of ibogaine and noribogaine obtained for individual subjects after single oral dose administrations demonstrate complex pharmacokinetic profiles. Ibogaine has shown preliminary efficacy for opiate detoxification and for short-term stabilization of drug-dependent persons as they prepare to enter substance abuse treatment. We report here that ibogaine significantly decreased craving for cocaine and heroin during inpatient detoxification. Self-reports of depressive symptoms were also significantly lower after ibogaine treatment and at 30 days after program discharge. Because ibogaine is cleared rapidly from the blood, the beneficial aftereffects of the drug on craving and depressed mood may be related to the effects of noribogaine on the central nervous system.

  8. A preliminary safety evaluation of polyhexamethylene guanidine hydrochloride.

    Science.gov (United States)

    Asiedu-Gyekye, Isaac Julius; Mahmood, Seidu Abdulai; Awortwe, Charles; Nyarko, Alexander Kwadwo

    2014-01-01

    Polyhexamethylene guanidine hydrochloride (PHMGH) is used worldwide as an antimicrobial agent with broad spectra of activity and also for treating pool water. This non-GLP preliminary study aims at investigating in a subchronic toxicity study possible effects at supra-optimal doses of this biocide. Both acute and subchronic toxicity studies were conducted. LD(50) for PHMGH was estimated to be 600 mg/kg (ie LC(50) 2 ml of 7.5% solution) when administered as a single dose by gavage via a stomach tube in accordance with the expected route of administration. The acute studies showed that the median lethal dose (LD(50)) of 600 mg/kg was accompanied by signs of neurotoxicity. Haematological and biochemical parameters of subchronic toxicity studies were non-significant. Subchronic doses of 0.006 mg/kg, 0.012 mg/kg and 0.036 mg/kg were administered. 20% of the animals at a dose of 0.006 mg/kg and 0.036 mg/kg showed mild degrees of hydropic changes in proximal tubules while 10% of animals at all the doses had their liver tissues showing local areas of mild pericentral hepatocytes degeneration. PHMGH did not produce any major organ defect with regard to the kidney, heart, and liver. The LD(50) was much higher than the recommended dosage by a factor of about 50,000. The recommended residual concentration is far less than the median lethal dose using rats as test subjects. These results could serve as a basis for investigating the full toxicological profile if it is to be used for the treatment of raw water to make it potable. © The Author(s) 2014.

  9. Preliminary study for unified management of CANDU safety codes and construction of database system

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, Hyoung Tae

    2003-03-01

    It is needed to develop the Graphical User Interface(GUI) for the unified management of CANDU safety codes and to construct database system for the validation of safety codes, for which the preliminary study is done in the first stage of the present work. The input and output structures and data flow of CATHENA and PRESCON2 are investigated and the interaction of the variables between CATHENA and PRESCON2 are identified. Furthermore, PC versions of CATHENA and PRESCON2 codes are developed for the interaction of these codes and GUI(Graphic User Interface). The PC versions are assessed by comparing the calculation results with those by HP workstation or from FSAR(Final Safety Analysis Report). Preliminary study on the GUI for the safety codes in the unified management system are done. The sample of GUI programming is demonstrated preliminarily. Visual C++ is selected as the programming language for the development of GUI system. The data for Wolsong plants, reactor core, and thermal-hydraulic experiments executed in the inside and outside of the country, are collected and classified following the structure of the database system, of which two types are considered for the final web-based database system. The preliminary GUI programming for database system is demonstrated, which is updated in the future work

  10. Preliminary safety evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)

    International Nuclear Information System (INIS)

    Dunn, T.D.; Lommers, L.J.; Tangirala, V.E.

    1994-04-01

    A qualitative comparison between the safety characteristics of the Gas Turbine-Modular Helium Reactor (GT-MHR) and those of the steam cycle shows that the two designs achieve equivalent levels of overall safety performance. This comparison is obtained by applying the scaling laws to detailed steam-cycle computations as well as the conclusions obtained from preliminary GT-MHR model simulations. The gas turbine design is predicted to be superior for some event categories, while the steam cycle design is better for others. From a safety perspective, the GT-MHR has a modest advantage for pressurized conduction cooldown events. Recent computational simulations of 102 column, 550 MW(t) GT-MHR during a depressurized conduction cooldown show that peak fuel temperatures are within the limits. The GT-MHR has a significantly lower risk due to water ingress events under operating conditions. Two additional scenarios, namely loss of load event and turbine deblading event that are specific to the GT-MHR design are discussed. Preliminary evaluation of the GT-MHR's safety characteristics indicate that the GT-MHR can be expected to satisfy or exceed its safety requirements

  11. Deep Brain Stimulation in Huntington’s Disease—Preliminary Evidence on Pathophysiology, Efficacy and Safety

    Directory of Open Access Journals (Sweden)

    Lars Wojtecki

    2016-08-01

    Full Text Available Huntington’s disease (HD is one of the most disabling degenerative movement disorders, as it not only affects the motor system but also leads to cognitive disabilities and psychiatric symptoms. Deep brain stimulation (DBS of the pallidum is a promising symptomatic treatment targeting the core motor symptom: chorea. This article gives an overview of preliminary evidence on pathophysiology, safety and efficacy of DBS in HD.

  12. Gas cooled fast reactor 2400 MWTh, status on the conceptual design studies and preliminary safety analysis

    International Nuclear Information System (INIS)

    Malo, J.Y.; Alpy, N.; Bentivoglio, F.

    2009-01-01

    The Gas cooled Fast Reactor (GFR) is considered by the French Commissariat a l'Energie Atomique as a promising concept, combining the benefits of fast spectrum and high temperature, using Helium as coolant. A status on the GFR preliminary viability was made at the end of 2007, ending the pre-conceptual design phase. A consistent overall systems arrangement was proposed and a preliminary safety analysis based on operating transient calculations and a simplified PSA had established a global confidence in the feasibility and safety of this baseline concept. Its potential for attractive performances had been pointed out. Compare to the more mature Sodium Fast Reactor technology, no demonstrator has ever been built and the feasibility demonstration will required a longer lead time. The next main project milestone is related to the GFR viability, scheduled in 2012. The current studies consist in revisiting the reactor reference design options as selected at the end of 2007. Most of them are being consolidated by going more in depth in the analysis. Some possible alternatives are assessed. The paper will give a status on the last studies performed on the core design and corresponding neutronics and cycle performance, the Decay Heat Removal strategy and preliminary safety analysis, systems design and balance of plant... This paper is complementary to the Icapp'09 papers 9062 dealing with the Gas cooled Fast Reactor Demonstrator ALLEGRO and 9378 related to GFR transients analysis. (author)

  13. Preliminary report on safety aspects on nuclear power generation in Sri Lanka

    International Nuclear Information System (INIS)

    Jayamanne, D.; Fernando, W.L.W.; Ariyadasa

    1988-01-01

    This document is intended as background information on nuclear energy to contribute to Sri Lanka's comparative study of alternative sources of energy. This study has considered the safety and environmental effects of nuclear power reactors. Basic concepts of nuclear physics are introduced and providing and appreciation of safety considerations and safety aspects of nuclear power plants and the personnel. Radioactive waste management, storage and disposal are also discussed. Natural radiation levels in Sri Lanka are provided as well as information on biological effects of radiation especially occupational exposure licensing procedures for nuclear power plants are outlined strategy for public awareness of nuclear power is proposed

  14. Conclusion of the Preliminary Safety report for the LILW Repository on Trgovska Gora

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.; Schaller, A.; Kucar-Dragicevic, S.; Cerskov Klika, M.; Subasic, D.

    2002-01-01

    For more than a decade, APO d.o.o. has been engaged in preparations which might lead to establishment of a radioactive waste repository on Trgovska Gora, suitable for disposal of low and intermediate level waste (LILW) from the nuclear power plant Krsko. A recent product of theses activities is the preliminary safety assessment report (PSAR) for the proposed repository. In addition to an extensive overview of the repository project status, this preliminary SAR describes how the safety assessment methodology is used to demonstrate that a LILW facility will comply with radiological protection and safety requirements after the repository closure. LILW repository is designed to isolate waste from the environment for a couple hundred years in a reasonably efficient manner. It is generally not practicable to grant full waste containment throughout that period, because it suffices to demonstrate that radionuclide release and migration will remain below acceptable levels, which is achieved through safety assessment scenarios, modeling and calculations. However, with very limited repository specific data, safety assessment can only produce a conservative estimate of the upper bounds of potential exposures the repository could inflict. This PSAR arrives at such estimates in two different ways: (a) by simple bounding calculations and (b) through more sophisticated modeling and application of dedicated computer codes, but with similar conservative assumptions. Both approaches conservatively estimate that the highest potential dose to a nearby resident cannot significantly exceed the dose constraint of 0.2 mSv per year. Only in case of inadvertent intrusion into the near-surface disposal vault, much higher doses might be inflicted immediately after the planned institutional control of 250 years expires, but that can be prevented by a longer control period. Despite the preliminary and bounding style of the calculations, the PSAR has identified most important assumptions and

  15. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.

  16. Light-water reactors: preliminary safety and environmental information document. Volume I

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle

  17. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  18. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers

  19. MICE - Absorber and focus coil safety working group design document: Preliminary design and assessments

    International Nuclear Information System (INIS)

    Barr, Giles; Baynham, Elwyn; Black, Edgar; Bradshaw, Tom; Cummings, Mary Anne; Green, Michael A.; Ishimoto, Shigeru; Ivanyushenkov, Yury; Lau, Wing; Zisman, Michael

    2003-01-01

    A Neutrino Factory based on a muon storage ring is the ultimate tool for studies of neutrino oscillations, including possibly the discovery of leptonic CP violation. it is also the first step toward a muon collider. To develop a stored-muon-beam facility to serve as a Neutrino Factory, it is necessary to ''cool'' a muon beam (decrease its phase-space volume). The short lifetime of the muon, 2.2 (micro)s at rest, eliminates all currently demonstrated cooling techniques and requires that a new, heretofore untried, technique--ionization cooling--be employed. Although ionization cooling of muons has never been demonstrated in practice, it has been shown by end-to-end simulation and design studies to be an important factor both for the performance and for the cost of a Neutrino Factory. This motivates an international program of R and D, including an experimental demonstration at Rutherford Appleton Laboratory (RAL). The aims of the international Muon Ionization Cooling Experiment are: (1) to show that it is possible to design, engineer and build a section of cooling channel capable of giving the desired performance for a Neutrino Factory; and (2) to place it in a muon beam and measure its performance in various modes of operation and beam conditions, thereby investigating the limits and practicality of cooling. The MICE collaboration has designed an experiment in which a section of an ionization cooling channel is exposed to a muon beam. This cooling channel assembles liquid-hydrogen absorbers providing energy loss and high-gradient radio frequency (RF) cavities to re-accelerate the particles, all tightly contained in a magnetic channel. It reduces the beam transverse emittance by > 10% for muon momenta between 140 and 240 MeV/c. The layout of the experiment is shown. They utilize one complete magnetic cell of the cooling channel, comprising three absorber-focus-coil (AFC) modules and two RF-coupling-coil (RFCC) modules. Spectrometers placed before and after the cooling section shown, perform the measurements of beam transmission and emittance reduction with an absolute precision of ± 0.1%. The detector solenoids each have separate coils (matching coils) to tailor the optics smoothly between the cooling channel and detector modules

  20. Mapping the nomological network of employee self-determined safety motivation: A preliminary measure in China.

    Science.gov (United States)

    Jiang, Li; Tetrick, Lois E

    2016-09-01

    The present study introduced a preliminary measure of employee safety motivation based on the definition of self-determination theory from Fleming (2012) research and validated the structure of self-determined safety motivation (SDSM) by surveying 375 employees in a Chinese high-risk organization. First, confirmatory factor analysis (CFA) was used to examine the factor structure of SDSM, and indices of five-factor model CFA met the requirements. Second, a nomological network was examined to provide evidence of the construct validity of SDSM. Beyond construct validity, the analysis also produced some interesting results concerning the relationship between leadership antecedents and safety motivation, and between safety motivation and safety behavior. Autonomous motivation was positively related to transformational leadership, negatively related to abusive supervision, and positively related to safety behavior. Controlled motivation with the exception of introjected regulation was negatively related to transformational leadership, positively related to abusive supervision, and negatively related to safety behavior. The unique role of introjected regulation and future research based on self-determination theory were discussed. Copyright © 2016 Elsevier Ltd. All rights reserved.

  1. Documenting Quality Improvement and Patient Safety Efforts: The Quality Portfolio. A Statement from the Academic Hospitalist Taskforce

    OpenAIRE

    Taylor, Benjamin B.; Parekh, Vikas; Estrada, Carlos A.; Schleyer, Anneliese; Sharpe, Bradley

    2013-01-01

    Physicians increasingly investigate, work, and teach to improve the quality of care and safety of care delivery. The Society of General Internal Medicine Academic Hospitalist Task Force sought to develop a practical tool, the quality portfolio, to systematically document quality and safety achievements. The quality portfolio was vetted with internal and external stakeholders including national leaders in academic medicine. The portfolio was refined for implementation to include an outlined fr...

  2. Supporting documents for LLL area 27 (410 area) safety analysis reports, Nevada Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Odell, B. N. [comp.

    1977-02-01

    The following appendices are common to the LLL Safety Analysis Reports Nevada Test Site and are included here as supporting documents to those reports: Environmental Monitoring Report for the Nevada Test Site and Other Test Areas Used for Underground Nuclear Detonations, U. S. Environmental Protection Agency, Las Vegas, Rept. EMSL-LV-539-4 (1976); Selected Census Information Around the Nevada Test Site, U. S. Environmental Protection Agency, Las Vegas, Rept. NERC-LV-539-8 (1973); W. J. Hannon and H. L. McKague, An Examination of the Geology and Seismology Associated with Area 410 at the Nevada Test Site, Lawrence Livermore Laboratory, Livermore, Rept. UCRL-51830 (1975); K. R. Peterson, Diffusion Climatology for Hypothetical Accidents in Area 410 of the Nevada Test Site, Lawrence Livermore Laboratory, Livermore, Rept. UCRL-52074 (1976); J. R. McDonald, J. E. Minor, and K. C. Mehta, Development of a Design Basis Tornado and Structural Design Criteria for the Nevada Test Site, Nevada, Lawrence Livermore Laboratory, Livermore, Rept. UCRL-13668 (1975); A. E. Stevenson, Impact Tests of Wind-Borne Wooden Missiles, Sandia Laboratories, Tonopah, Rept. SAND 76-0407 (1976); and Hydrology of the 410 Area (Area 27) at the Nevada Test Site.

  3. 1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1977-08-01

    This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations

  4. HACCP and water safety plans in Icelandic water supply: preliminary evaluation of experience.

    Science.gov (United States)

    Gunnarsdóttir, María J; Gissurarson, Loftur R

    2008-09-01

    Icelandic waterworks first began implementing hazard analysis and critical control points (HACCP) as a preventive approach for water safety management in 1997. Since then implementation has been ongoing and currently about 68% of the Icelandic population enjoy drinking water from waterworks with a water safety plan based on HACCP. Preliminary evaluation of the success of HACCP implementation was undertaken in association with some of the waterworks that had implemented HACCP. The evaluation revealed that compliance with drinking water quality standards improved considerably following the implementation of HACCP. In response to their findings, waterworks implemented a large number of corrective actions to improve water safety. The study revealed some limitations for some, but not all, waterworks in relation to inadequate external and internal auditing and a lack of oversight by health authorities. Future studies should entail a more comprehensive study of the experience with the use of HACCP with the purpose of developing tools to promote continuing success.

  5. 1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    1977-08-01

    This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations. (DLC)

  6. Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    1997-01-01

    In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

  7. European passive plant program preliminary safety analyses to support system design

    International Nuclear Information System (INIS)

    Saiu, Gianfranco; Barucca, Luciana; King, K.J.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

  8. Preliminary safety assessment study for the conceptual design of a repository in tuff at Yucca Mountain

    International Nuclear Information System (INIS)

    Jackson, J.L.; Gram, H.F.; Hong, K.J.; Ng, H.S.; Pendergrass, A.M.

    1984-12-01

    Preliminary estimates of the upper bounds on postulated worst-case radiological releases resulting from possible accidents during the operating period of a prospective repository in tuff at Yucca Mountain are presented. Possible disrupting events are screened to identify the accidents of greatest potential consequence. The radiological dose commitments for the general public and repository personnel are estimated for postulated releases caused by natural phenomena, man-made events, and operational accidents. All postulated worst-case releases result in doses to the public that are lower than the 0.5-rem, whole-body dose-per-accident limit set by the Nuclear Regulatory Commission (NRC) in 10 CFR 60. Doses to repository personnel are within the NRC's 5.0-rem/yr occupational exposure limit set in 10 CFR 20 for normal operations. Doses are within this limit for all accidents except the transportation accident and fire in a drift. A preliminary risk assessment has also been performed. Based on this preliminary safety study, the proposed site boundaries and design criteria routinely used in constructing nuclear facilities appear to be adequate to protect the safety of the general public during the operating phase of the repository

  9. Safety administration division business report. The second quarter of 2002. Document on present state of affairs

    International Nuclear Information System (INIS)

    Ishibashi, Takashi

    2003-02-01

    The activities of Safety Administration Division covers many fields in Tokai-Works such as the management of a labor safety health, the crisis management and the security, the safeguards of the nuclear materials, the transport of nuclear materials, and the management of a quality assurance. This report is summary of the activities of Safety Administration Division in July to September in 2002. (author)

  10. Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

    International Nuclear Information System (INIS)

    Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

    2007-01-01

    SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

  11. Geoscientific long-term prognosis. Preliminary safety analysis for the site Gorleben

    International Nuclear Information System (INIS)

    Mrugalla, Sabine

    2011-07-01

    The preliminary safety analysis of the site Gorleben includes the following chapters: (1) Introduction; (2) Aim and content of the geoscientific long-term prognosis for the site Gorleben; (3) Boundary conditions at the site Gorleben: climate; geomorphology; overlying rocks and adjoining rocks; hydrogeology; salt deposit Gorleben. (4) Probable future geological developments at the site Gorleben: supraregional developments with effects on the site Gorleben; glacial period developments; developments of the geomorphology, overlying and adjoining rocks; future developments of the hydrological systems at the site Gorleben; future saliniferous specific developments of the salt deposit Gorleben. (5) Commentary on the unlikely or excludable developments of the site Gorleben.

  12. Operating performance and environmental and safety risks: A preliminary comparison of majors and independents

    International Nuclear Information System (INIS)

    Pulsipher, A.G.; Iledare, W.O.; Baumann, R.H.; Mesyanzhinov, D.

    1995-01-01

    The objective is to compare the safety and environmental records of oil and gas companies operating on the OCS in the Gulf of Mexico over the past decade. The reason for doing so is to help inform public sector policy-makers and private sector decision-makers about the potential safety and environmental risks associated with the expected increased presence of smaller independents in the domestic oil and gas industry in general and on the federal OCS in particular. The preliminary conclusion is that although independents have had a modestly high incidence of fires and explosions than the majors, the difference is not significant statistically and is largely attributable to a few ''bad actors'' rather than demonstrably poorer practice by the group as a whole

  13. 14 CFR Appendix B of Part 415 - Safety Review Document Outline

    Science.gov (United States)

    2010-01-01

    ... Performance Graphs 2.0Launch Operator Organization (§ 415.111) 2.1Launch Operator Organization (§ 415.111 and... Plan 4.3.1Flight Safety Personnel 4.3.2Flight Safety Rules 4.3.3Flight Safety System Summary and... Instrumentation Plan 6.2Configuration Management and Control Plan 6.3Frequency Management Plan 6.4Flight...

  14. Customer Relationship Management System in Occupational Safety & Health Companies: Research on Practice and Preliminary Design Solution

    Directory of Open Access Journals (Sweden)

    Robert Fabac

    2011-10-01

    Full Text Available One of the most prominent contemporary trends in formation of companies is the approach to development of a customer-oriented company. In this matter, various versions related to the intensity of this orientation are differentiated. Customer relationship management (CRM system is a well-known concept, and its practice is being studied and improved in connection to various sectors. Companies providing services of occupational safety and health (OHS mainly cooperate with a large number of customers and the quality of this cooperation largely affects the occupational safety and health of employees. Therefore, it is of both scientific and wider social interest to study and improve the relationship of these companies with their customers. This paper investigates the practice of applying CRM in Croatian OHS companies. It identifies the existing conditions and suggests possible improvements in the practice of CRM, based on experts’ assessments using analytic hierarchy process evaluation. Universal preliminary design was created as a framework concept for the formation of a typical customer-oriented OHS services company. Preliminary design includes a structural view, which provides more details through system diagrams, and an illustration of main cooperation processes of a company with its customer.

  15. Ares-I-X Vehicle Preliminary Range Safety Malfunction Turn Analysis

    Science.gov (United States)

    Beaty, James R.; Starr, Brett R.; Gowan, John W., Jr.

    2008-01-01

    Ares-I-X is the designation given to the flight test version of the Ares-I rocket (also known as the Crew Launch Vehicle - CLV) being developed by NASA. As part of the preliminary flight plan approval process for the test vehicle, a range safety malfunction turn analysis was performed to support the launch area risk assessment and vehicle destruct criteria development processes. Several vehicle failure scenarios were identified which could cause the vehicle trajectory to deviate from its normal flight path, and the effects of these failures were evaluated with an Ares-I-X 6 degrees-of-freedom (6-DOF) digital simulation, using the Program to Optimize Simulated Trajectories Version 2 (POST2) simulation framework. The Ares-I-X simulation analysis provides output files containing vehicle state information, which are used by other risk assessment and vehicle debris trajectory simulation tools to determine the risk to personnel and facilities in the vicinity of the launch area at Kennedy Space Center (KSC), and to develop the vehicle destruct criteria used by the flight test range safety officer. The simulation analysis approach used for this study is described, including descriptions of the failure modes which were considered and the underlying assumptions and ground rules of the study, and preliminary results are presented, determined by analysis of the trajectory deviation of the failure cases, compared with the expected vehicle trajectory.

  16. UPSAT guidelines. 1996 edition. Reference document for IAEA Uranium Productions Safety Assessment Teams (UPSATs)

    International Nuclear Information System (INIS)

    1996-05-01

    The IAEA Uranium Production Safety Assessment Team (UPSAT) programme provides advice and assistance to Member States to enhance the safety and environmental performance of uranium production facilities during construction, commissioning and operation. Sound design and construction are prerequisite for the safe and environmentally responsible operation of uranium mines and mills. However, the safety of the facility depends ultimately on sound policies, procedures and practices; on the capability and reliability of the construction, commissioning and operating personnel; on comprehensive instructions; and on adequate resources. A positive attitude and conscientiousness on the part of the management and staff in discharging their responsibilities is important to safety. The UPSAT guidelines have been developed in the following areas: (1) management, organization and administration; (2) training and qualification; (3) operation (4) maintenance; (5) safety, fire protection, emergency planning, and preparedness; (6) radiation protection; (7) environmental monitoring programme; (8) construction management; (9) commissioning and decommissioning

  17. Preliminary safety evaluation of an aircraft impact on a near-surface radioactive waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Lo Frano, R.; Forasassi, G.; Pugliese, G. [Department of Industrial and Civil Engineering (DICI), University of Pisa, Pisa (Italy)

    2013-07-01

    The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a 'beyond design basis' event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a 'robust' target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircraft, even in presence of ongoing concrete progressive failure (some penetration and spalling of the

  18. Preliminary standard review guide for Environmental Restoration/Decontamination and Decommissioning safety analyses

    International Nuclear Information System (INIS)

    Ellingson, D.R.

    1993-06-01

    The review guide is based on the shared experiences, approaches, and philosophies of the Environmental Restoration/Decontamination and Decommissioning (ER/D ampersand D) subgroup members. It is presented in the form of a review guide to maximize the benefit to both the safety analyses practitioner and reviewer. The guide focuses on those challenges that tend to be unique to ER/D ampersand D cleanup activities. Some of these experiences, approaches, and philosophies may find application or be beneficial to a broader spectrum of activities such as terminal cleanout or even new operations. Challenges unique to ER/D ampersand D activities include (1) consent agreements requiring activity startup on designated dates; (2) the increased uncertainty of specific hazards; and (3) the highly variable activities covered under the broad category of ER/D ampersand D. These unique challenges are in addition to the challenges encountered in all activities; e.g., new and changing requirements and multiple interpretations. The experiences in approaches, methods, and solutions to the challenges are documented from the practitioner and reviewer's perspective, thereby providing the viewpoints on why a direction was taken and the concerns expressed. Site cleanup consent agreements with predetermined dates for restoration activity startup add the dimension of imposed punitive actions for failure to meet the date. Approval of the safety analysis is a prerequisite to startup. Actions that increase expediency are (1) assuring activity safety; (2) documenting that assurance; and (3) acquiring the necessary approvals. These actions increase the timeliness of startup and decrease the potential for punitive action. Improvement in expediency has been achieved by using safety analysis techniques to provide input to the line management decision process rather than as a review of line management decisions. Expediency is also improved by sharing the safety input and resultant decisions with

  19. International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

    International Nuclear Information System (INIS)

    2016-02-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  20. Developing a disaster education program for community safety and resilience: The preliminary phase

    Science.gov (United States)

    Nifa, Faizatul Akmar Abdul; Abbas, Sharima Ruwaida; Lin, Chong Khai; Othman, Siti Norezam

    2017-10-01

    Resilience encompasses both the principles of preparedness and reaction within the dynamic systems and focuses responses on bridging the gap between pre-disaster activities and post-disaster intervention and among structural/non-structural mitigation. Central to this concept is the ability of the affected communities to recover their livelihood and inculcating necessary safety practices during the disaster and after the disaster strikes. While these ability and practices are important to improve the community safety and resilience, such factors will not be effective unless the awareness is present among the community. There have been studies conducted highlighting the role of education in providing awareness for disaster safety and resilience from a very young age. However for Malaysia, these area of research has not been fully explored and developed based on the specific situational and geographical factors of high-risk flood disaster locations. This paper explores the importance of disaster education program in Malaysia and develops into preliminary research project which primary aim is to design a flood disaster education pilot program in Kampung Karangan Primary School, Kelantan, Malaysia.

  1. Safety

    International Nuclear Information System (INIS)

    2001-01-01

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

  2. Los Alamos National Laboratory corregated metal pipe saw facility preliminary safety analysis report. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-09-19

    This Preliminary Safety Analysis Report addresses site assessment, facility design and construction, and design operation of the processing systems in the Corrugated Metal Pipe Saw Facility with respect to normal and abnormal conditions. Potential hazards are identified, credible accidents relative to the operation of the facility and the process systems are analyzed, and the consequences of postulated accidents are presented. The risk associated with normal operations, abnormal operations, and natural phenomena are analyzed. The accident analysis presented shows that the impact of the facility will be acceptable for all foreseeable normal and abnormal conditions of operation. Specifically, under normal conditions the facility will have impacts within the limits posted by applicable DOE guidelines, and in accident conditions the facility will similarly meet or exceed the requirements of all applicable standards. 16 figs., 6 tabs.

  3. Safety administration division business report. The fourth quarter of 2001. Document on present state of affairs

    International Nuclear Information System (INIS)

    Kanamori, Masashi

    2002-06-01

    The business of the Safety administration Division became a wide range such as the management of a labor safety health, the crisis management, the security and the management of an entrance, and the business of the following concerning the Tokai Works, the protection of nuclear materials, the business of the sanction, the nuclear material safeguards, the transport of nuclear materials and the business of a quality assurance. For the purpose of summarizing these businesses and utilizing the data concerning the businesses, the report about the businesses achievement has been periodically drawn up as quarter news since 2001, when the Safety Administration Division was established. This report describes about the business achievement of the fourth quarter news from January to March in 2002. (author)

  4. Safety administration division business report. The third quarter of 2001. Document on present state of affairs

    International Nuclear Information System (INIS)

    Kanamori, Masashi

    2002-04-01

    The business of the Safety administration Division became a wide range such as the management of a labor safety health, the crisis management, the security and the management of an entrance, and the business of the following concerning the Tokai Works, the protection of nuclear materials, the business of the sanction, the nuclear material safeguards, the transport of nuclear materials and the business of a quality assurance. For the purpose of summarizing these business and utilizing the data concerning the business, the report about the businesses achievement has been periodically drawn up as quarter news since 2001, when the Safety Administration Division was established. This report describes about the business achievement of the third quarter news from October to December in 2001. (author)

  5. Preliminary safety criteria for organic watch list tanks at the Hanford site

    International Nuclear Information System (INIS)

    Webb, A.B.; Stewart, J.L.; Turner, O.A.; Plys, M.G.; Malinovic, B.; Grigsby, J.M.; Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J.

    1995-11-01

    Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended

  6. Preliminary safety criteria for organic watch list tanks at the Hanford site

    Energy Technology Data Exchange (ETDEWEB)

    Webb, A.B.; Stewart, J.L.; Turner, O.A. [Westinghouse Hanford Co., Richland, WA (United States); Plys, M.G.; Malinovic, B. [Fauske and Associates, Inc., Burr Ridge, IL (United States); Grigsby, J.M. [G & P Consulting, Inc. (United States); Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J. [Pacific Northwest Lab., Portland, OR (United States)

    1995-11-01

    Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended.

  7. Preliminary Assessment for the Effects of the External Hazard Factors on the Safety of NPPs

    International Nuclear Information System (INIS)

    Jin, So Beom; Hyun, Seung Gyu; Kim, Sang Yun; Lee, Sung Kyu; Hur, Youl

    2010-01-01

    The Ch.etsu Offshore Earthquake(2007.7.16) in Japan caused damage to the Kashiwazaki-Kariwa(K-K) Nuclear Power Plants (NPPs) with seismic ground motion that exceeded the design level. This incident drew the interest of the safety evaluation studies for NPPs subjected to earthquakes exceeding the design basis around the world. Also, the Indian Ocean Tsunami(2004.12.26) tripped the Madras NPP by reason of flooding of the intake pump house and inundated the construction site of a fast breeder reactor site in India. In addition, from the various man-made and natural hazards such as the oil spill accident near Mallipo, Taean, Chungnam (2007.12), the forest fire near the Ulchin NPP site, the several inflows of marine organism into the intake of the Ulchin NPP, it was confirmed that the safety of NPPs may be affected by natural and human induced disasters. Intergovernmental Panel on Climate Change (IPCC) has been warned about global warming; the average temperature rose about 1.5 .deg. C during the 20th century and the damages caused by typhoons and heavy rains have also increased in Korea. Accordingly, a natural disaster prevention research team(hereafter team) ,which have been organized and operated since 2009, has assessed the impact of various hazards such as earthquakes and environmental changes due to global warming on the safety of NPP and has discussed to establish countermeasures. This paper introduces that the preliminary assessment for the effects of the external hazard factors on the safety of NPPs was conducted by the team

  8. Spent nuclear fuel project-criteria document Cold Vacuum Drying Facility phase 2 safety analysis report

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1998-01-01

    The criteria document provides the criteria and guidance for developing the SNF CVDF Phase 2 SAR. This SAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, and installation acceptance testing of the CVDF systems

  9. Preliminary investigation of interconnected systems interactions for the safety injection system of Indian Point-3

    International Nuclear Information System (INIS)

    Alesso, H.P.; Lappa, D.A.; Smith, C.F.; Sacks, I.J.

    1983-01-01

    The rich diversity of ideas and techniques for analyzing interconnected systems interaction has presented the NRC with the problem of identifying methods appropriate for their own review and audit. This report presents the findings of a preliminary study using the Digraph Matrix Analysis method to evaluate interconnected systems interactions for the safety injection system of Indian Point-3. The analysis effort in this study was subjected to NRC constraints regarding the use of Boolean logic, the construction of simplified plant representations or maps, and the development of heuristic measures as specified by the NRC. The map and heuristic measures were found to be an unsuccessful approach. However, from the effort to model and analyze the Indian Point-3 safety injection system, including Boolean logic in the model, singleton and doubleton cut-sets were identified. It is recommended that efforts excluding Boolean logic and utilizing the NRC heuristic measures not be pursed further and that the Digraph Matrix approach (or other comparable risk assessment technique) with Boolean logic included to conduct the audit of the Indian Point-3 systems interaction study

  10. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  11. Significance of the existing normative and technical documentation to promote ecologic safety

    OpenAIRE

    Fatima Ermakhanova

    2010-01-01

    The author justifies the need to develop new standards for gas purification process. It is believed that these standards should assume introduction of new resource-saving technologies and also meet the modern international requirements to improve environmental safety. The article shows the need for the introduction and development of environmental management system according to ISO 14000 standards to reduce the industrial impact of gas deposits on the environment.

  12. Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

    International Nuclear Information System (INIS)

    1999-11-01

    A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as 36 Cl and 93 Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon

  13. Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-01

    A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as {sup 36}Cl and {sup 93}Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon.

  14. Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Teague, Jonathan Gayle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-30

    Criticality safety at TA-55 relies on nuclear material containers that are water resistant to prevent significant amounts of water from coming into contact with fissile material in the event of a fire that causes a breach of glovevbox confinement and subsequent fire water ingress. A “water tight container” is a container that will not allow more than 50ml of water ingress when fully submerged, except when under sufficient pressure to produce structural discontinuity. There are many types of containers, welded containers, hermetically sealed containers, filtered containers, etc.

  15. DSRS guidelines. Reference document for the IAEA Design Safety Review Services

    International Nuclear Information System (INIS)

    1999-01-01

    The publication covers the general topic of design safety review of a nuclear power plant. It is intended to make Member States aware of the possibility of a service through which they can have a better appreciation of the overall design of a facility or of a plant already in operation. It includes a generic and procedural part followed by a technical part corresponding to different systems of a nuclear power plant. It is intended to be used mainly in preparation and execution of a design review service by the IAEA and to provide information to potential recipients of the service regarding the effort involved and the topics that can be covered. it is expected to be useful if Member States decide to conduct such reviews themselves either through regulatory authorities or as part of self assessment activities by plant management

  16. Preliminary safety equipment list for Tank 241-C-106 Manipulator Retrieval System, Project W-340

    International Nuclear Information System (INIS)

    Guthrie, R.L.

    1994-01-01

    This document identifies the anticipated safety classification of the estimated major subsystems, based on the projected major functions, that will be used as guidance for the development of the conceptual design of the Manipulator Retrieval System for Tank 241-C-106. This document is intended to be updated as the design of the Manipulator Retrieval System evolves through the conceptual and definitive design phases. The Manipulator Retrieval System is to be capable of removing the hardened sludge heel at the bottom of single shell Tank 241-C-106 and to perform an overall clean out of the tank that leaves a maximum of 360 ft 3 (TPA milestone M-45-00). The thickness of the heel prior to initiation of waste retrieval with the Manipulator Retrieval System is estimated to be 1- to 2-ft. The Manipulator Retrieval System is currently in the pre-conceptual phase with no definitive systems or subsystems. The anticipated retrieval functions for the Manipulator Retrieval System is based on Table 6-2 of WHC-SD-W340-ES-001, Rev. 1. Projected equipment to accomplish these functions were based on the following systems and equipment: Rotary Mode Core Sampling Equipment (WHC-SD-WM-SEL-032); Light Duty Utility Arm System Equipment (WHC-SD-WM-SEL-034); Single Shell Tanks Equipment (WHC-SD-WM-SEL-020)

  17. School Survey on Crime and Safety (SSOCS) 2000 Public-Use Data Files, User's Manual, and Detailed Data Documentation. [CD-ROM].

    Science.gov (United States)

    National Center for Education Statistics (ED), Washington, DC.

    This CD-ROM contains the raw, public-use data from the 2000 School Survey on Crime and Safety (SSOCS) along with a User's Manual and Detailed Data Documentation. The data are provided in SAS, SPSS, STATA, and ASCII formats. The User's Manual and the Detailed Data Documentation are provided as .pdf files. (Author)

  18. Safety assessment document (SAD) for the Princeton Beta Experiment Modification (PBX-M)

    International Nuclear Information System (INIS)

    Stencel, J.R.; Parsells, R.F.

    1988-04-01

    The Princeton Beta Experiment-Modification (PBX-M) is an experimental device of the tokamak type. A tokamak is characterized by a strong toroidal magnetic field composed of an externally driven component parallel to the torus centerline modified by the field produced by a transformer-driven current (OH) in the confined plasma. A second magnetic field parallel to the major toroidal axis is added to provide radial equilibrium for the plasma. As an advanced tokamak, PBX-M will have additional magnetic fields to reshape the plasma cross section from a circle into a kidney bean shape; it will also be equipped with 6MW or more of auxiliary heating power provided by four neutral beam injectors, with RF systems, and with an extensive set of diagnostics. Potential hazards associated with PBX-M, which are analyzed in this report, result from energy stored in the magnetic fields, high voltages necessary for the operation of some of the equipment and diagnostics, neutron radiation when the neutral beams are run with deuterium and x-rays, especially those emitted as a result of plasma-wall interaction. This report satisfies the requirements set forth in the PPPL Health and Safety Directives, specifically HSD-5003, and in DOE Order 5481.1B and its Chicago operations supplement (DOE86, DOE82)

  19. New Reactor Siting in Finland, Hanhikivi Site in Pyhaejoki - STUK preliminary safety assessment

    International Nuclear Information System (INIS)

    Nevalainen, Janne

    2013-01-01

    STUK has performed a preliminary assessment of the Decision-in-Principle on the Fennovoima application. A variety of factors must be considered in the selection of a site, including effects of the site on the plant design and the effects of the plant on the site environment. These include external hazards, both natural and human-induced. Since this is a new site, an extensive siting process is followed, that can include an EIA. A site survey is performed to identify candidate sites, after investigating a large region and rejecting unsuitable sites. The remaining sites are then screened and compared on the basis of safety and other considerations to select one or more preferred sites. Natural hazards include geology, seismology, hydrology and meteorology. Offshore ice will be a particular hazard for this plant, since the site is on average only 1.5 m above sea level. The design basis earthquake corresponds to a return frequency of 100,000 years, with 50 % confidence. The existing sites in southern Finland used a design peak ground acceleration of 0.1 g with the ground response spectrum maximum at 10 Hz. The candidate sites in northern Finland will require a peak ground acceleration of 0.2 g with the ground response spectrum maximum at 25 Hz

  20. Safety assessment document for the environmental test complex (Building 834) at Site 300

    International Nuclear Information System (INIS)

    Odell, B.N.; Pfeifer, H.E.

    1981-01-01

    A safety assessment was performed to determine if accidents occurring at the 834 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (HE) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire. Seismic and extreme wind (including missiles) analyses indicate that the buildings are basically sound. (However, there are a few recommendations to further enhance the structural integrity of these facilities). Additional lightning protection for these facilities is being installed. These buildings are located high above the dry creek bed so that a flood is improbable. A criticality or a high explosive detonation involving plutonium is very remote since the radioactive materials are encased and plutonium and HE are not permitted concurrently in the same area at Site 300. (The exceptions to this policy are that explosive actuating devices are sometimes located in assemblies containing fissile materials. However, a planned or accidental actuation will not effect the safe containment of the fissile material within the assembly). Even though the possibility of an HE explosion involving uranium and beryllium is remote, the off-site lung doses were calculated and found to be below the accepted standards. It was determined that a fire was unlikely due to the low fire loading and the absence of ignition sources. It was also determined that the consequences of any accidents were reduced by the remote location of these facilities, their design, and by administrative controls

  1. Safety and feasibility of transcranial direct current stimulation in pediatric hemiparesis: randomized controlled preliminary study.

    Science.gov (United States)

    Gillick, Bernadette T; Feyma, Tim; Menk, Jeremiah; Usset, Michelle; Vaith, Amy; Wood, Teddi Jean; Worthington, Rebecca; Krach, Linda E

    2015-03-01

    Transcranial direct current stimulation (tDCS) is a form of noninvasive brain stimulation that has shown improved adult stroke outcomes. Applying tDCS in children with congenital hemiparesis has not yet been explored. The primary objective of this study was to explore the safety and feasibility of single-session tDCS through an adverse events profile and symptom assessment within a double-blind, randomized placebo-controlled preliminary study in children with congenital hemiparesis. A secondary objective was to assess the stability of hand and cognitive function. A double-blind, randomized placebo-controlled pretest/posttest/follow-up study was conducted. The study was conducted in a university pediatric research laboratory. Thirteen children, ages 7 to 18 years, with congenital hemiparesis participated. Adverse events/safety assessment and hand function were measured. Participants were randomly assigned to either an intervention group or a control group, with safety and functional assessments at pretest, at posttest on the same day, and at a 1-week follow-up session. An intervention of 10 minutes of 0.7 mA tDCS was applied to bilateral primary motor cortices. The tDCS intervention was considered safe if there was no individual decline of 25% or group decline of 2 standard deviations for motor evoked potentials (MEPs) and behavioral data and no report of adverse events. No major adverse events were found, including no seizures. Two participants did not complete the study due to lack of MEP and discomfort. For the 11 participants who completed the study, group differences in MEPs and behavioral data did not exceed 2 standard deviations in those who received the tDCS (n=5) and those in the control group (n=6). The study was completed without the need for stopping per medical monitor and biostatisticial analysis. A limitation of the study was the small sample size, with data available for 11 participants. Based on the results of this study, tDCS appears to be safe

  2. Preliminary safety evaluation for the Laxemar subarea. Based on data and site descriptions after the initial site investigation stage

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-03-15

    The main objectives of this Preliminary Safety Evaluation (PSE) of the Laxemar subarea have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB in 2000. These criteria both concern properties of the site judged to be necessary for safety and engineering (requirements) and properties judged to be beneficial (preferences). The findings are then evaluated in order to provide feedback to continued investigations and design work. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The latter is eventually done in coming Safety Assessments. This preliminary safety evaluation shows that, according to existing data, the Laxemar subarea meets all safety requirements. The evaluation also shows that the Laxemar subarea meets most of the safety preferences, but for some aspects of the site description further reduction of the uncertainties would enhance the safety case. Despite the stated concerns, there is no reason, from a safety point of view, not to continue the Site Investigations at the Laxemar subarea. There are uncertainties to resolve and the safety would eventually need to be verified through a proper safety assessment. Only some of the uncertainties noted in the Site Descriptive Model have safety implications and need further resolution for this reason. Furthermore, uncertainties may need resolving for other reasons, such as giving an adequate assurance of site understanding or assisting in optimising design. Notably, there are questions about the

  3. Preliminary safety evaluation for the Laxemar subarea. Based on data and site descriptions after the initial site investigation stage

    International Nuclear Information System (INIS)

    Andersson, Johan

    2006-03-01

    The main objectives of this Preliminary Safety Evaluation (PSE) of the Laxemar subarea have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB in 2000. These criteria both concern properties of the site judged to be necessary for safety and engineering (requirements) and properties judged to be beneficial (preferences). The findings are then evaluated in order to provide feedback to continued investigations and design work. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The latter is eventually done in coming Safety Assessments. This preliminary safety evaluation shows that, according to existing data, the Laxemar subarea meets all safety requirements. The evaluation also shows that the Laxemar subarea meets most of the safety preferences, but for some aspects of the site description further reduction of the uncertainties would enhance the safety case. Despite the stated concerns, there is no reason, from a safety point of view, not to continue the Site Investigations at the Laxemar subarea. There are uncertainties to resolve and the safety would eventually need to be verified through a proper safety assessment. Only some of the uncertainties noted in the Site Descriptive Model have safety implications and need further resolution for this reason. Furthermore, uncertainties may need resolving for other reasons, such as giving an adequate assurance of site understanding or assisting in optimising design. Notably, there are questions about the

  4. Computerising documentation

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The nuclear power generation industry is faced with public concern and government pressures over safety, efficiency and risk. Operators throughout the industry are addressing these issues with the aid of a new technology - technical document management systems (TDMS). Used for strategic and tactical advantage, the systems enable users to scan, archive, retrieve, store, edit, distribute worldwide and manage the huge volume of documentation (paper drawings, CAD data and film-based information) generated in building, maintaining and ensuring safety in the UK's power plants. The power generation industry has recognized that the management and modification of operation critical information is vital to the safety and efficiency of its power plants. Regulatory pressure from the Nuclear Installations Inspectorate (NII) to operate within strict safety margins or lose Site Licences has prompted the need for accurate, up-to-data documentation. A document capture and management retrieval system provides a powerful cost-effective solution, giving rapid access to documentation in a tightly controlled environment. The computerisation of documents and plans is discussed in this article. (Author)

  5. Safety of high speed ground transportation systems: X2000 US demonstration vehicle dynamics trials, preliminary test report. Report for October 1992-January 1993

    Energy Technology Data Exchange (ETDEWEB)

    Whitten, B.T.; Kesler, J.K.

    1993-01-01

    The report documents the procedures, events, and results of vehicle dynamic tests carried out on the ASEA-Brown Boveri (ABB) X2000 tilt body trainset in the US between October 1992 and January 1993. These tests, sponsored by Amtrak and supported by the FRA, were conducted to assess the suitability of the X2000 trainset for safe operation at elevated cant deficiencies and speeds in Amtrak's Northeast Corridor under existing track conditions in a revenue service demonstration. The report describes the safety criteria against which the performance of the X2000 test train was examined, the instrumentation used, the test locations, and the track conditions. Preliminary results are presented from tests conducted on Amtrak lines between Philadelphia and Harrisburg, PA, and between Washington DC and New York NY, in which cant deficiencies of 12.5 inches and speeds of 154 mph were reached in a safe and controlled manner. The significance of the results is discussed, and preliminary conclusions and recommendations are presented.

  6. Preliminary safety analysis of high-intensity interval training (HIIT) in persons with chronic stroke.

    Science.gov (United States)

    Carl, Daniel L; Boyne, Pierce; Rockwell, Bradley; Gerson, Myron; Khoury, Jane; Kissela, Brett; Dunning, Kari

    2017-03-01

    The purpose of this study was to assess safety via electrocardiographic (ECG), blood pressure (BP), heart rate (HR), and orthopedic responses to 3 different high-intensity interval training (HIIT) protocols in persons with stroke. Eighteen participants (10 male; 61.9 + 8.3 years of age; 5.8 + 4.2 years poststroke) completed a symptom-limited graded exercise test (GXT) with ECG monitoring to screen for eligibility and determine HR peak. The 3 HIIT protocols involved repeated 30 s bursts of treadmill walking at maximum speed alternated with rest periods of 30 s (P30), 1 min (P60), or 2 min (P120). Sessions were performed in random order and included 5 min warm up, 20 min HIIT, and 5 min cool down. Variables measured included ECG activity, BP, HR, signs and symptoms of cardiovascular intolerance, and orthopedic concerns. Generalized linear mixed models and Tukey-Kramer adjustment were used to compare protocols using p HIIT session. HIIT elicited HRs in excess of 88% of measured HR peak including 6 (P30), 8 (P60), and 2 (P120) participants eliciting a HR response above their GXT HR peak . Both maximum BP and HR were significantly higher in P30 and P60 relative to P120. Preliminary data indicate that persons with chronic stroke who have been prescreened with an ECG stress test, a symptom-limited GXT, and a harness for fall protection may safely participate in HIIT, generating substantially higher HRs than what is seen in traditional moderate intensity training.

  7. Proposals for the Radioactive Substances (Basic Safety Standards) (England and Wales) Regulations 2000 and the Radioactive Substances (Basic Safety Standards) (England and Wales) Direction 2000. Consultative document

    International Nuclear Information System (INIS)

    2000-01-01

    This document contains proposals for changes to the Radioactive Substances Act 1993 (RSA 93) and proposals for a Direction to be given to the Environment Agency in order to implement aspects of the European Directive 96/29/Euratom concerned with the control of radioactive waste. The Directive lays down basic safety standards for the protection of the health of workers and the general public against the dangers arising from ionising radiation. With the Government pledged to making government more accessible and responsive, an important feature of this approach is effective consultation with all interested organisations. This leads to more realistic and robust proposals, which is particularly important when dealing with proposed legislation. In March this year, the Government published a consultation paper 'The Radioactive Substances Act 1993: Implementing the Revised Basic Safety Standards Directive Euratom 96/29.' This sought comments on the basic principles for change - including the setting of levels of radioactivity below which radioactive material should be considered outside the framework of regulatory control. This document forms the second stage of the consultation process with the aim of gathering views on the proposed legal instruments to implement the Directive. This document: explains the background to the proposed regulations (paragraphs 8-13); summarises the results of the consultation on principles (paragraphs 14-24); describes the proposed changes (paragraphs 25-36); includes draft Regulations (paragraphs 27-29); includes a draft Direction to the Environment Agency (paragraphs 30-36); describes the next steps (paragraphs 37-39); includes a draft Regulatory Impact Assessment (paragraphs 40-41). In general, the devolved administrations in Scotland, Wales and Northern Ireland have assumed responsibility for environmental issues and hence management of radioactive waste policies and legislation affecting their respective countries. However, this

  8. Posiva's application for a decision in principle concerning a disposal facility for spent nuclear fuel. STUK's statement and preliminary safety appraisal

    International Nuclear Information System (INIS)

    Ruokola, E.

    2000-03-01

    In May 1999, Posiva Ltd submitted to the Government an application, pursuant to the Nuclear Energy Act, for a Decision in Principle on a disposal facility for spent nuclear fuel from the Finnish nuclear power plants. The Ministry of Trade and Industry requested the Radiation and Nuclear Safety Authority (STUK) to draw up a preliminary safety appraisal concerning the proposed disposal facility. In the beginning of this report, STUK's statement to the Ministry and Industry concerning the proposed disposal facility is given. In that statement, STUK concludes that the Decision in Principle is currently justified from the standpoint of safety. The statement is followed by a safety appraisal, where STUK deems, how the proposed disposal concept, site and facility comply with the safety requirements included in the Government's Decision (478/1999). STUK's preliminary safety appraisal was supported by contributions from a number of outside experts. A collective opinion by an international group of ten distinguished experts is appended to this report. (orig.)

  9. Joint SKI and SSI review of SKB preliminary safety assessment of repository for long-lived low- and intermediate-level waste. Review report

    International Nuclear Information System (INIS)

    2001-03-01

    included no discussion in the safety report as to which R and D activities they intend to prioritise. According to the current SKB timetable, siting and construction of SFL 3-5 will not begin for another 30 years. However, SKI and SSI do not consider this to be a reason to postpone essential R and D work. If a complete and thorough basis is not produced for assessing the long-term safety of an SFL 3-5 repository, the risk that these waste categories will have to undergo interim storage for an indefinite period of time increases. A future siting of SFL 3-5 based on our current level of knowledge is problematic. The present safety assessment points toward a substantial site-specific effect on the repository's protective capacity that can be related primarily to the local groundwater flow rate, but also to relevant geochemical conditions. Calculated doses for cases involving consumption of drinking water give the impression that the margins are small vis-a-vis the existing requirement framework, at least based on the methods used heretofore. In their main report, SKB discuss the possibility of improving the technical barriers to increase their impact on long-term safety (thereby mitigating the impact of site-specific factors). SKI and SSI feel that this approach is reasonable from the current preliminary perspective, but not for subsequent stages. SKB should in future formulate a proposed repository design that can be considered sufficiently robust with respect to the effects of the site-specific factors and their long-term evolution. The requirements and criteria that are relevant to the siting of SFL 3-5 must be addressed therein. In addition, more in-depth studies regarding the optimum storage depth for SFL 3-5 and the importance of the interactions between SFL 2 and SFL 3-5 should be undertaken relatively soon. The importance of these issues needs to be well documented in order to provide a basis for identifying suitable rock volumes for potential siting of SFL 3-5. Once

  10. Preliminary study on functional performance of compound type multistage safety injection tank

    International Nuclear Information System (INIS)

    Bae, Youngmin; Kim, Young In; Kim, Keung Koo

    2015-01-01

    Highlights: • Functional performance of compound type multistage safety injection tanks is studied. • Effects of key design parameters are scrutinized. • Distinctive flow features in compound type safety injection tanks are explored. - Abstract: A parametric study is carried out to evaluate the functional performance of a compound type multistage safety injection tank that would be considered one of the components for the passive safety injection systems in nuclear power plants. The effects of key design parameters such as the initial volume fraction and charging pressure of gas, tank elevation, vertical location of a sparger, resistance coefficient, and operating condition on the injection flow rate are scrutinized along with a discussion of the relevant flow features. The obtained results indicate that the compound type multistage safety injection tank can effectively control the injection flow rate in a passive manner, by switching the driving force for the safety injection from gas pressure to gravity during the refill and reflood phases, respectively

  11. A preliminary study on the design in architecture of nuclear and radiation safety standard system

    International Nuclear Information System (INIS)

    Song Dahu; Zhang Chi; Yang Lili; Li Bin; Liu Yingwei; An Hongzhen; Gao Siyi; Liu Ting; Meng De

    2014-01-01

    The connotation and function of nuclear and radiation safety standards are analyzed, and their relationships with the relevant laws and regulations are discussed in the paper. Some suggestions and blue print of overall architecture to build nuclear and radiation safety standard system are proposed, on the basis of researching the application status quo, existing problems and needs for nuclear and radiation safety standards in China. This work is a beneficial exploration and attempt to establish China's nuclear and radiation safety standards. (authors)

  12. Implementation in Russia and the European Union of International Safety Standards of Identity Documents with Biometric Data: Legal Regulation and Perspectives

    Directory of Open Access Journals (Sweden)

    Alexander Grigoryevich Volevodz

    2015-01-01

    Full Text Available The article contains the findings of a research into particular aspects of use of identity documents with personal biometric data. It considers the international safety standards of documents with biometric data worked out by the International Civil Aviation Organization (ICAO, pursuant to which those data should be included into machine-readable documents used by their holders for travel to various states. It contains the information on the implementation of these international standards in Russian and European Union law. The author has substantiated a conclusion to the effect that the procedure established in Russia for production and issuance, as well as for use of international, diplomatic and service passports identifying the Russian Federation citizen outside the Russian Federation territory, containing electronic information carriers with personal and biometric personal data, currently conforms to the international safety standards of documents with biometric data. The article surveys the experience of introducing domestic biometric identity documents - electronic passports in various countries of the world, and the problems arising therefrom. It substantiates the advantages and disadvantages of determining a passport of the Russian Federation citizen issued in the form of an identity card with an electronic information carrier, as the main document of the Russian Federation citizen identifying him domestically within the country's territory.

  13. Quarterly report on the activities in safety administration division. The third quarter of 2002. Document on present state of affairs

    International Nuclear Information System (INIS)

    Ishibashi, Takashi

    2003-06-01

    The activities of Safety Administration Division covers many fields in Tokai-Works such as the management of a labor safety health, the crisis management and the security, the safeguards of the nuclear materials, the transport of nuclear materials, and the management of a quality assurance. This report is summary of the activities of Safety Administration Division in October to December in 2002. (author)

  14. 77 FR 50720 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2012-08-22

    ... Used in Safety Systems of Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft... Computer Software used in Safety Systems of Nuclear Power Plants.'' The DG-1207 is proposed Revision 1 of... for Digital Computer Software Used in Safety Systems of Nuclear Power Plants'' is temporarily...

  15. Feasibility, safety, acceptability, and preliminary efficacy of measurement-based care depression treatment for HIV patients in Bamenda, Cameroon.

    Science.gov (United States)

    Pence, Brian W; Gaynes, Bradley N; Atashili, Julius; O'Donnell, Julie K; Kats, Dmitry; Whetten, Kathryn; Njamnshi, Alfred K; Mbu, Tabenyang; Kefie, Charles; Asanji, Shantal; Ndumbe, Peter

    2014-06-01

    Depression affects 18-30 % of HIV-infected patients in Africa and is associated with greater stigma, lower antiretroviral adherence, and faster disease progression. However, the region's health system capacity to effectively identify and treat depression is limited. Task-shifting models may help address this large mental health treatment gap. Measurement-Based Care (MBC) is a task-shifting model in which a Depression Care Manager guides a non-psychiatric (e.g., HIV) provider in prescribing and managing antidepressant treatment. We adapted MBC for depressed HIV-infected patients in Cameroon and completed a pilot study to assess feasibility, safety, acceptability, and preliminary efficacy. We enrolled 55 participants; all started amitriptyline 25-50 mg daily at baseline. By 12 weeks, most remained at 50 mg daily (range 25-125 mg). Median (interquartile range) PHQ-9 depressive severity scores declined from 13 (12-16) (baseline) to 2 (0-3) (week 12); 87 % achieved depression remission (PHQ-9 feasibility, safety, acceptability, and preliminary efficacy in this uncontrolled pilot study. Further research should assess whether MBC could improve adherence and HIV outcomes in this setting.

  16. Preliminary post-closure safety assessment of repository concepts for low level radioactive waste at the Bruce Site, Ontario

    International Nuclear Information System (INIS)

    Little, R.H.; Penfold, J.S.S.; Egan, M.J.; Leung, H.

    2005-01-01

    The preliminary post-closure safety assessment of permanent repository concepts for low-level radioactive waste (LLW) at the Ontario Power Generation (OPG) Bruce Site is described. The study considered the disposal of both short and long-lived LLW. Four geotechnically feasible repository concepts were considered (two near-surface and two deep repositories). An approach consistent with best international practice was used to provide a reasoned and comprehensive analysis of post-closure impacts of the repository concepts. The results demonstrated that the deep repository concepts in shale and in limestone, and the surface repository concept on sand should meet radiological protection criteria. For the surface repository concept on glacial till, it appears that increased engineering such as grouting of waste and voids should be considered to meet the relevant dose constraint. Should the project to develop a permanent repository for LLW proceed, it is expected that this preliminary safety assessment would need to be updated to take account of future site-specific investigations and design updates. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  18. Safety evaluation report related to the preliminary design of the Standard Reference System, RESAR-414

    International Nuclear Information System (INIS)

    1978-11-01

    The safety evaluation for the Westinghouse Standard Reactor includes information on general reactor characteristics; design criteria for systems and components; reactor coolant system; engineered safety systems; instrumentation and controls; electric power systems; auxiliary systems; steam and power conversion system; radioactive waste management; radiation protection; conduct of operations; accident analyses; and quality assurance

  19. A preliminary study on safety stock placement in capacitated supply chains

    NARCIS (Netherlands)

    Sitompul, Carles; Aghezzaf, El Houssaine; Chen, Huan; Dullaert, Wout

    2006-01-01

    The issue of safety stock placement is one of the challenging problems in the area of supply chain design. Safety stocks play a significant role in supply chains since they assure the service level and the smoothness of the flow of materials throughout the chain. Some special models of the problem

  20. A PRELIMINARY STUDY ON SAFETY STOCK PLACEMENT IN CAPACITATED SUPPLY CHAINS

    NARCIS (Netherlands)

    Sitompul, Carles; Aghezzaf, El Houssaine; Chen, Huan; Dullaert, Wout; Dolgui, A.; Morel, G.; Pereira, C.E.

    2006-01-01

    Abstract The issue of safety stock placement is one of the challenging problems in the area of supply chain design. Safety stocks play a significant role in supply chains since they assure the service level and the smoothness of the flow of materials throughout the chain. Some special models of the

  1. Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    2013-01-01

    The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

  2. Documents preparation and review

    International Nuclear Information System (INIS)

    1999-01-01

    Ignalina Safety Analysis Group takes active role in assisting regulatory body VATESI to prepare various regulatory documents and reviewing safety reports and other documentation presented by Ignalina NPP in the process of licensing of unit 1. The list of main documents prepared and reviewed is presented

  3. Criteria for safety-related nuclear plant operator actions: a preliminary assessment of available data

    International Nuclear Information System (INIS)

    Haas, P.M.; Bott, T.F.

    1980-01-01

    In the US, an effort has been underway for a number of years to develop a design standard to define when required manual operator action can be accepted as part of a nuclear plant design basis. Insufficient data are available to provide quantitative guidelines for the standard. To provide the necessary data base to support such standards and the necessary quantitative assessment of operator reliability, the US Nuclear Regulatory Commission is sponsoring a study at Oak National Laboratory to develop the data base. A preliminary assessment completed in April, 1979 concluded that sufficient data from US operating experience did not exist to provide an adequate data base. A program of research using full-scope nuclear plant simulators and results that are correlated to field data was suggested. That program was recently initiated. The approach, results and conclusions of the preliminary assessment are reviewed and the planned research program of simulator studies is summarised. (author)

  4. Preliminary Assessment of Operational Hazards and Safety Requirements for Airborne Trajectory Management (ABTM) Roadmap Applications

    Science.gov (United States)

    Cotton, William B.; Hilb, Robert; Koczo, Stefan, Jr.; Wing, David J.

    2016-01-01

    A set of five developmental steps building from the NASA TASAR (Traffic Aware Strategic Aircrew Requests) concept are described, each providing incrementally more efficiency and capacity benefits to airspace system users and service providers, culminating in a Full Airborne Trajectory Management capability. For each of these steps, the incremental Operational Hazards and Safety Requirements are identified for later use in future formal safety assessments intended to lead to certification and operational approval of the equipment and the associated procedures. Two established safety assessment methodologies that are compliant with the FAA's Safety Management System were used leading to Failure Effects Classifications (FEC) for each of the steps. The most likely FEC for the first three steps, Basic TASAR, Digital TASAR, and 4D TASAR, is "No effect". For step four, Strategic Airborne Trajectory Management, the likely FEC is "Minor". For Full Airborne Trajectory Management (Step 5), the most likely FEC is "Major".

  5. Preliminary safety concept for disposal of the very low level radioactive waste in Romania.

    Science.gov (United States)

    Niculae, O; Andrei, V; Ionita, G; Duliu, O G

    2009-05-01

    In Romania, there are certain nuclear installations in operation or under decommissioning, all of them representing an important source of very low level waste (VLLW). This paper presents an overview on the approach of the VLLW management in Romania, focused on those resulted from the nuclear power plants decommissioning. At the same time, the basic elements of safety concept, together with some safety evaluations concerning VLLW repository are presented and discussed too.

  6. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

  7. Occupational health and safety in the Moroccan construction sites: preliminary diagnosis

    Science.gov (United States)

    Tarik, Bakeli; Adil, Hafidi Alaoui

    2018-05-01

    Managing occupational health and safety on Moroccan construction sector represents the first step for projects' success. In fact, by avoiding accidents, all the related direct and indirect costs and delays can be prevented. That leads to an important question always asked by any project manager: what are the factors responsible for accidents? How can they be avoided? Through this research, the aim is to go through these questions, to contribute in occupational health and safety principles understanding, to identify construction accidentology and risk management opportunities and to approach the case of Moroccan construction sites by an accurate diagnosis. The approach is to make researchers, managers, stakeholders and deciders aware about the criticality of construction sites health and safety situation. And, to do the first step for a scientific research project in relation with health and safety in the Moroccan construction sector. For this, the paper will study the related state of art namely about construction sites accidents causation, and will focus on Reason's `Swiss cheese' model and its utilization for Moroccan construction sites health and safety diagnosis. The research will end with an estimation of an accidents fatality rate in the Moroccan construction sector and a benchmarking with the international rates. Finally, conclusions will be presented about the necessity of Occupational Health and Safety Management System (OHSMS) implementation, which shall cover all risk levels, and insure, at the same time, that the necessary defenses against accidents are on place.

  8. The management of electronic documents generated from compilation and revision processes of nuclear and radiation safety regulations and standards

    International Nuclear Information System (INIS)

    Wang Wenhai; Fan Yun; Shang Zhaorong

    2010-01-01

    As the Secretary Group of Regulations and Standards Review Committee on nuclear and radiation safe needs to deal with a large number of electronic documents in course of the regulation and standard review meetings, the article gives a systematical method including electronic document file naming and management as well as procedures of file transfer, storage and usage. (authors)

  9. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    International Nuclear Information System (INIS)

    Tyagi, Himanshu; Soni, Jignesh; Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli; Gahlaut, Agrajit; Joshi, Jaydeep; Parmar, Deepak; Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun

    2016-01-01

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  10. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    Energy Technology Data Exchange (ETDEWEB)

    Tyagi, Himanshu, E-mail: htyagi@iter-india.org [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Soni, Jignesh [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Gahlaut, Agrajit [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Joshi, Jaydeep; Parmar, Deepak [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India)

    2016-11-15

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  12. 78 FR 47805 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2013-08-06

    ... Used in Safety Systems of Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Revision... Used in Safety Systems of Nuclear Power Plants.'' This RG endorses the Institute of Electrical and... nuclear power plants. ADDRESSES: Please refer to Docket ID NRC-2012-0195 when contacting the NRC about the...

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

    International Nuclear Information System (INIS)

    Moon, Kwangsu; Kim, Sa Kil; Oh, Yeon Ju; Shin, Youmin; Lee, Yong-Hee; Jang, Tong Il

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  15. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

    2017-03-15

    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  16. Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

    International Nuclear Information System (INIS)

    1985-04-01

    The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated

  17. Preliminary safety evaluation for the Simpevarp subarea. Based on data and site descriptions after the initial site investigation stage

    International Nuclear Information System (INIS)

    2005-04-01

    The main objectives of this Preliminary safety evaluation (PSE) of the Simpevarp subarea are: to determine, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB in the report SKB-TR--00-12. These criteria both concern properties of the site judged to be necessary for safety and engineering (requirements) and properties judged to be beneficial (preferences). The findings are then evaluated in order to provide feedback to continued investigations and design work. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The evaluation shows that even considering remaining uncertainties, the Simpevarp subarea meets all safety requirements and most of the safety preferences. Consequently, from a safety point of view, there is no reason not to continue the Site Investigations of the Simpevarp subarea. There are still uncertainties to resolve and the safety would eventually need to be verified through a full safety assessment. Still, this Preliminary Safety Evaluation demonstrates that it is likely that a safe repository for spent nuclear fuel of the KBS-3 type could be constructed at the site. The following feedback is provided to the site investigations and the associated site modelling: Reducing the uncertainty on the deformation zone geometry within the Simpevarp subarea would allow for a more specified layout, although the sensitivity analysis shows that the space needed is rather robust with respect to uncertainties in the zones. There is substantial uncertainty in the discrete fracture network (DFN) model

  18. Criteria for safety related nuclear plant operator actions: a preliminary assessment of available data

    International Nuclear Information System (INIS)

    Haas, P.M.; Bott, T.F.

    1982-01-01

    The need for a quantitative data base on the reliability of nuclear power plant operators has long been recognised by human factors and reliability analysts, and the great need for further assessment of operator performance under accident conditions has been dramatically emphasised by the incident at Three Mile Island-2. In the US, an effort has been under way for a number of years to develop a design standard to define when required manual operator action can be accepted as part of a nuclear plant design basis. To provide the necessary data base to support such standards and the necessary quantitative assessment of operator reliability, the US Nuclear Regulatory Commission is sponsoring a study at Oak Ridge National Laboratory to develop the data base. A preliminary assessment, completed in April 1979, concluded that sufficient data from US operating experience did not exist to provide an adequate data base. A programme of research using full-scope nuclear plant simulators and results that are correlated to field data was suggested. That programme was recently initiated. This paper reviews the approach, results and conclusions of the preliminary assessment and summarises the planned research programme of simulator studies. (author)

  19. Model review and evaluation for application in DOE safety basis documentation of chemical accidents - modeling guidance for atmospheric dispersion and consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Woodarad, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanna, S. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hesse, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, J. -C. [Argonne National Lab. (ANL), Argonne, IL (United States); Lewis, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mazzola, C. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    1997-09-01

    The U.S. Department of Energy (DOE), through its Defense Programs (DP), Office of Engineering and Operations Suppon, established the Accident Phenomenology and Consequence (AP AC) Methodology Evaluation Program to identify and evaluate methodologies and computer codes to support accident phenomenological and consequence calculations for both radiological and nonradiological materials at DOE facilities and to identify development needs. The program is also intended to define and recommend "best or good engineering/safety analysis practices" to be followed in preparing ''design or beyond design basis" assessments to be included in DOE nuclear and nonnuclear facility safety documents. The AP AC effort is intended to provide scientifically sound and more consistent analytical approaches, by identifying model selection procedures and application methodologies, in order to enhance safety analysis activities throughout the DOE complex.

  20. Preliminary Safety Analysis of the Gorleben Site: Safety Concept and Application to Scenario Development Based on a Site-Specific Features, Events and Processes (FEP) Database - 13304

    Energy Technology Data Exchange (ETDEWEB)

    Moenig, Joerg; Beuth, Thomas; Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Theodor-Heuss-Str. 4, D-38122 Braunschweig (Germany); Lommerzheim, Andre [DBE TECHNOLOGY GmbH, Eschenstr. 55, D-31224 Peine (Germany); Mrugalla, Sabine [Federal Institute for Geosciences and Natural Resources (BGR), Stilleweg 2, D-30655 Hannover (Germany)

    2013-07-01

    Based upon the German safety criteria, released in 2010 by the Federal Ministry of the Environment (BMU), a safety concept and a safety assessment concept for the disposal of heat-generating high-level waste have both been developed in the framework of the preliminary safety case for the Gorleben site (Project VSG). The main objective of the disposal is to contain the radioactive waste inside a defined rock zone, which is called containment-providing rock zone. The radionuclides shall remain essentially at the emplacement site, and at the most, a small defined quantity of material shall be able to leave this rock zone. This shall be accomplished by the geological barrier and a technical barrier system, which is required to seal the inevitable penetration of the geological barrier by the construction of the mine. The safe containment has to be demonstrated for probable and less probable evolutions of the site, while evolutions with very low probability (less than 1 % over the demonstration period of 1 million years) need not to be considered. Owing to the uncertainty in predicting the real evolution of the site, plausible scenarios have been derived in a systematic manner. Therefore, a comprehensive site-specific features, events and processes (FEP) data base for the Gorleben site has been developed. The safety concept was directly taken into account, e.g. by identification of FEP with direct influence on the barriers that provide the containment. No effort was spared to identify the interactions of the FEP, their probabilities of occurrence, and their characteristics (values). The information stored in the data base provided the basis for the development of scenarios. The scenario development methodology is based on FEP related to an impairment of the functionality of a subset of barriers, called initial barriers. By taking these FEP into account in their probable characteristics the reference scenario is derived. Thus, the reference scenario describes a

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

    International Nuclear Information System (INIS)

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

    1983-06-01

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

  2. Preliminary safety examination on thorium molten-salt nuclear energy synergetics

    International Nuclear Information System (INIS)

    Furukawa, Kazuo

    1991-01-01

    The new global fission industry for the next century should keep a strong public acceptance, which means to ensure an enough rational safety feature not only in the engineering issue but also in all issues of integral fuel-cycle system. In this sense, the safety characteristics of the Thorium Molten-Salt Nuclear Energy Synergetic System (THORIMS-NES) is widely examined relating with the several aspects of environmental (including resources, radio-waste, etc.) social (including anti-nuclear proliferation and terrorism, etc), basic technological, engineering, institutional, and economical aspects. From this examination it seems that this system is verified as one of the most promising measures of North-South problem, greenhouse effect, etc in the world. (author). 11 refs., 3 figs., 5 tabs

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

  4. A preliminary CATHENA thermalhydraulic model of the Canadian SCWR for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wang, D.F.; Wang, S., E-mail: wangd@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-06-15

    The supercritical water-cooled reactor (SCWR) is one of six reactor concepts under development in the Generation-IV International Forum (GIF). As a member of GIF, Canada is developing a pressure-tube type SCWR, which has the potential to fulfill all major GIF goals on enhanced safety, sustainability, economics, and proliferation resistance. The system thermalhydraulics code CATHENA will be used in the safety analyses for the Canadian SCWR. Based on the current conceptual design of the Canadian SCWR, a CATHENA idealization has been developed. This model includes all 336 fuel channels with a detailed model of heat transfer in the reactor core. Also modeled are the main pumps, inlet plenum, outlet plenum, turbines, and heavy water moderator. In this paper, the CATHENA idealization of the Canadian SCWR conceptual design is described. Simulation results for steady-state normal operations are also presented for the current Canadian SCWR conceptual design. (author)

  5. Health and safety: Preliminary comparative assessment of the Satellite Power System (SPS) and other energy alternatives

    Science.gov (United States)

    Habegger, L. J.; Gasper, J. R.; Brown, C.

    1980-01-01

    Data readily available from the literature were used to make an initial comparison of the health and safety risks of a fission power system with fuel reprocessing; a combined-cycle coal power system with a low-Btu gasifier and open-cycle gas turbine; a central-station, terrestrial, solar photovoltaic power system; the satellite power system; and a first-generation fusion system. The assessment approach consists of the identification of health and safety issues in each phase of the energy cycle from raw material extraction through electrical generation, waste disposal, and system deactivation; quantitative or qualitative evaluation of impact severity; and the rating of each issue with regard to known or potential impact level and level of uncertainty.

  6. Health and safety. Preliminary comparative assessment of the satellite power system (SPS) and other energy alternatives

    International Nuclear Information System (INIS)

    Habegger, L.J.; Gasper, J.R.; Brown, C.D.

    1980-04-01

    Existing data on the health and safety risks of a satellite power system and four electrical generation systems are analyzed: a combined-cycle coal power system with a low-Btu gasifier and open-cycle gas turbine, a fission power system with fuel reprocessing, a central-station, terrestrial, solar-photovoltaic power system, and a first-generation design for a fusion power system. The systems are compared on the basis of expected deaths and person-days lost per year associated with 1000 MW of average electricity generation and the number of health and safety risks that are identified as potentially significant but unquantifiable. The appendices provide more detailed information on risks, uncertainties, additional research needed, and references for the identified impacts of each system

  7. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  8. White paper: Preliminary assessment of LNG vehicle technology, economics, and safety issues (Revision 1). Topical report, April-August 1991

    International Nuclear Information System (INIS)

    Powars, C.; Lucher, D.; Moyer, C.; Browning, L.

    1992-01-01

    The objective of the study is to evaluate the potential of LNG as a vehicle fuel, to determine market niches, and to identify needed technology improvements. The white paper is being issued when the work is approximately 30 percent complete to preview the study direction, draw preliminary conclusions, and make initial recommendations. Interim findings relative to LNG vehicle technology, economics, and safety are presented. It is important to decide if heavier hydrocarbons should be allowed in LNG vehicle fuel. Development of suitable refueling couplings and vehicle fuel supply pressure systems are recommended. Initial economics analyses considered transit buses and pickup and delivery trucks fueled via onsite liquefiers and imported LNG. Net user costs were more than (but in some cases close to) those for diesel fuel and gasoline. Lowering the cost of small-scale liquefiers would significantly improve the economics of LNG vehicles. New emissions regulations may introduce considerations beyond simple cost comparisons. LNG vehicle safety and available accident data are reviewed. Consistent codes for LNG vehicles and refueling facilities are needed

  9. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  10. Joint SKI and SSI review of SKB preliminary safety assessment of repository for long-lived low- and intermediate-level waste. Review report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    that SKB have included no discussion in the safety report as to which R and D activities they intend to prioritise. According to the current SKB timetable, siting and construction of SFL 3-5 will not begin for another 30 years. However, SKI and SSI do not consider this to be a reason to postpone essential R and D work. If a complete and thorough basis is not produced for assessing the long-term safety of an SFL 3-5 repository, the risk that these waste categories will have to undergo interim storage for an indefinite period of time increases. A future siting of SFL 3-5 based on our current level of knowledge is problematic. The present safety assessment points toward a substantial site-specific effect on the repository's protective capacity that can be related primarily to the local groundwater flow rate, but also to relevant geochemical conditions. Calculated doses for cases involving consumption of drinking water give the impression that the margins are small vis-a-vis the existing requirement framework, at least based on the methods used heretofore. In their main report, SKB discuss the possibility of improving the technical barriers to increase their impact on long-term safety (thereby mitigating the impact of site-specific factors). SKI and SSI feel that this approach is reasonable from the current preliminary perspective, but not for subsequent stages. SKB should in future formulate a proposed repository design that can be considered sufficiently robust with respect to the effects of the site-specific factors and their long-term evolution. The requirements and criteria that are relevant to the siting of SFL 3-5 must be addressed therein. In addition, more in-depth studies regarding the optimum storage depth for SFL 3-5 and the importance of the interactions between SFL 2 and SFL 3-5 should be undertaken relatively soon. The importance of these issues needs to be well documented in order to provide a basis for identifying suitable rock volumes for potential

  11. Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

    Energy Technology Data Exchange (ETDEWEB)

    Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2013-06-15

    Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

  12. Safety of operations in the manufacture of driver fuel for the first charge of the Dragon Reactor and modifications to the safety document for the Dragon Fuel Element Production Building

    International Nuclear Information System (INIS)

    Beutler, H.; Cross, J.; Flamm, J.

    1965-01-01

    The manufacture of the zirconium containing 'driver' fuel and fuel elements for the First Charge of the Dragon Reactor Experiment has been completed without incident. This is a report on the safety of operations in the Dragon Fuel Element Production Building during an approximately six month period when the 'driver' fuel was manufactured and 25 elements containing this fuel were assembled and exported to the Reactor Building. The opportunity is taken to bring the Safety Document up-to-date and to report on any significant operational failures of equipment. (author)

  13. The role the safety analysis report and other documents in the licensing procedure in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Albrecht, E.; Ballhaus, S.

    1975-01-01

    Legal requirements for the licensing procedure, requirements for and preparation of licensing documents, authorities and experts participating in the licensing procedure, course of the atomic licensing procedure e.g. application for construction and operation, public hearing, site and concept license, partial licenses. (HP) [de

  14. Research and development of technologies for safe and environmentally optimal recovery and disposal of explosive wastes. Task 2, Preliminary impact assessment for environment, health and safety (EIA)

    Energy Technology Data Exchange (ETDEWEB)

    Duijm, N.J.; Markert, F. [Risoe (Denmark); Larsen, S.G. [DEMEX A/S (Denmark)

    1998-09-01

    As described in the project proposal `Research and Development of Technologies for Safe and Environmentally optimal recovery and Disposal of Explosive Wastes`, dated 31. May 1996, the objective of Task 2, Preliminary Impact Assessment for Environment, Health and Safety, is to: Analyse the environmental impact of noise and emissions to air, water and soil; Assess the risk of hazards to workers` health and safety and to the public. Task 2, Preliminary Impact Assessment for Environment, Health and Safety (EIA), has been performed from August 1997 to September 1998. First, a methodology has been established, based on Multi-Criteria Decision Analysis (MCDA), to select the `best` technology on the basis of clearly defined objectives, including minimal impacts on environment, health and safety. This included a review of different types of explosive waste with a focus on the environment implications, identifying the issues relevant to defining the criteria or objectives with respect to environment and safety in the framework of explosive waste, as well as the preliminary definition of objectives for the final impact assessment. Second, the previously identified recovery and disposal technologies (Task 1) have been qualitatively assessed on the basis of the relevant objectives. This qualitative assessment includes also economic considerations and an attempt to rank the technologies in an MCDA framework. (au)

  15. Preliminary results from the application of risk matrix method for safety assessment in industrial radiography

    International Nuclear Information System (INIS)

    Lopez G, A.; Cruz, D.; Truppa, W.; Aravena, M.; Tamayo, B.

    2015-09-01

    Although the uses of ionizing radiation in industry are subject to procedures that provide a high level of safety, experience has shown that equipment failure, human errors, or the combination of both that can trigger accidental exposures may occur. Traditionally, the radiation safety checks whether these industrial practices (industrial radiography, industrial irradiators, among others) are sufficiently safe to prevent similar accidental exposures already occurred, so that becomes dependent on the published information and not always answers questions like: What other events can occur, or what other risks are present? Taking into account the results achieved by the Foro Iberoamericano de Organismos Reguladores Radiologicos y Nucleares, its leading position in the use of techniques of risk analysis in radioactive facilities and the need to develop a proactive approach to the prevention of accidents arising from the use of ionizing radiations in the industry, it intends to apply the risk analysis technique known as Risk Matrix to a hypothetical reference entity for the region in which industrial radiography is performed. In this paper the results of the first stage of this study are shown, that is the identification of initiating events (IE) and barriers that help mitigate the consequences of such IE, so that can appreciate the applicability of this method to industrial radiography services, to reduce the risk to acceptable levels. The fundamental advantage associated with the application of this methodology is that can be applied by the professionals working in the service and identifies specific weaknesses that from the point of view of safety there, so they can be prioritized resources depending on risk reduction. (Author)

  16. Preliminary safety assessment of C-8 xylitol monoester and xylitol phosphate esters.

    Science.gov (United States)

    Silveira, J E P S; Pereda, M C V; Nogueira, C; Dieamant, G; Cesar, C K M; Assanome, K M; Silva, M S; Torello, C O; Queiroz, M L S; Eberlin, S

    2016-02-01

    Most of the cosmetic compounds with preservative properties available in the market pose some risks concerning safety, such as the possibility of causing sensitization. Due to the fact that there are few options, the proper development of new molecules with this purpose is needed. Xylitol is a natural sugar, and the antimicrobial properties of xylitol-derived compounds have already been described in the literature. C-8 xylitol monoester and xylitol phosphate esters may be useful for the development of skincare products. As an initial screen for safety of chemicals, the combination of in silico methods and in vitro testing can aid in prioritizing resources in toxicological investigations while reducing the ethical and monetary costs that are related to animal and human testing. This study was designed to evaluate the safety of C-8 xylitol monoester and xylitol phosphate esters regarding carcinogenicity, mutagenicity, skin and eye irritation/corrosion and sensitization through alternative methods. For the initial safety assessment, quantitative structure-activity relationship methodology was used. The prediction of the parameters carcinogenicity/mutagenicity, skin and eye irritation/corrosion and sensitization was generated from the chemical structure. The analysis also comprised physical-chemical properties, Cramer rules, threshold of toxicological concern and Michael reaction. In silico results of candidate molecules were compared to 19 compounds with preservative properties that are available in the market. Additionally, in vitro tests (Ames test for mutagenicity, cytotoxicity and phototoxicity tests and hen's egg test--chorioallantoic membrane for irritation) were performed to complement the evaluation. In silico evaluation of both molecules presented no structural alerts related to eye and skin irritation, corrosion and sensitization, but some alerts for micronucleus and carcinogenicity were detected. However, by comparison, C-8 xylitol monoester, xylitol

  17. Swine plasma immunoglobulins for prevention and treatment of post-weaning diarrhoea: Safety and Preliminary results

    DEFF Research Database (Denmark)

    Hedegaard, Chris Juul; Strube, Mikael Lenz; Bendix Hansen, Marie

    . coli F4+ induced PWD, we observed that piglets given IgG as a feed supplement cleared the E coli infection significantly faster than control weaner piglets not receiving an immunoglobulin feed supplement. Furthermore, deep sequencing of the ileal microbiota showed a significantly lowered colonization...... their adhesion to porcine epithelial cells in vitro. As the immunoglobulin fraction is intended for oral use as a feed supplement, we also tested the safety of feeding 4 grams of natural immunoglobulins to 4-5 week old weaner piglets for 14 days and observed no adverse effects. In an experimental model of E...

  18. Strategic environmental safety inspection for the National disposal program. Description of the inspection volume. Documentation for the scoping team

    International Nuclear Information System (INIS)

    2015-01-01

    The Strategic environmental safety inspection for the National disposal program covers the following topics: Legal framework: determination of the requirement for an environmental inspection program, coordination of the scoping team into the overall context; environmental targets; approach for assessment and evaluation of environmental impact, description of the inspection targets for the strategic environmental inspection; consideration of alternatives.

  19. Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

    International Nuclear Information System (INIS)

    OSCAR, DEBBY S.; WALKER, SHARON ANN; HUNTER, REGINA LEE; WALKER, CHERYL A.

    1999-01-01

    The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2

  20. Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

    Energy Technology Data Exchange (ETDEWEB)

    OSCAR,DEBBY S.; WALKER,SHARON ANN; HUNTER,REGINA LEE; WALKER,CHERYL A.

    1999-12-01

    The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2.

  1. Preliminary safety analysis of the Baita Bihor radioactive waste repository, Romania

    International Nuclear Information System (INIS)

    Little, Richard; Bond, Alex; Watson, Sarah; Dragolici, Felicia; Matyasi, Ludovic; Matyasi, Sandor; Naum, Mihaela; Niculae, Ortenzia; Thorne, Mike

    2007-01-01

    A project funded under the European Commission's Phare Programme 2002 has undertaken an in-depth analysis of the operational and post-closure safety of the Baita Bihor repository. The repository has accepted low- and some intermediate-level radioactive waste from industry, medical establishments and research activities since 1985 and the current estimate is that disposals might continue for around another 20 to 35 years. The analysis of the operational and post-closure safety of the Baita Bihor repository was carried out in two iterations, with the second iteration resulting in reduced uncertainties, largely as a result taking into account new information on the hydrology and hydrogeology of the area, collected as part of the project. Impacts were evaluated for the maximum potential inventory that might be available for disposal to Baita Bihor for a number of operational and postclosure scenarios and associated conceptual models. The results showed that calculated impacts were below the relevant regulatory criteria. In light of the assessment, a number of recommendations relating to repository operation, optimisation of repository engineering and waste disposals, and environmental monitoring were made. (authors)

  2. Preliminary safety evaluation for the Forsmark area. Based on data and site descriptions after the initial site investigation stage

    International Nuclear Information System (INIS)

    Andersson, Johan

    2005-08-01

    The main objectives of this Preliminary Safety Evaluation (PSE) of the Forsmark area have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The evaluation shows that, even considering remaining uncertainties, the Forsmark area meets all stated safety requirements and preferences. Consequently, from a safety point of view, there is no reason not to continue the Site Investigations of the Forsmark area. There are still uncertainties to resolve and the safety would eventually need to be verified through a full safety assessment. Nevertheless, this Preliminary Safety Evaluation demonstrates that it is likely that a safe repository for spent nuclear fuel of the KBS-3 type could be constructed at the site. The following feedback is provided to the site investigations and the associated site modelling: Reducing the uncertainty on the deformation zone geometry inside the target area would be needed to more firmly define locations of the suitable deposition volumes. There is substantial uncertainty in the Discrete Fracture Network model. Further reduction of the uncertainties, if needed, would probably only be possible from the underground, detailed investigation phase. Efforts need also be spent on improving the DFN-modelling. There are assumptions made in current models that could be challenged and there seems to be room for better use of the borehole information. It is particularly important to provide

  3. Preliminary safety evaluation for the Forsmark area. Based on data and site descriptions after the initial site investigation stage

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-08-01

    The main objectives of this Preliminary Safety Evaluation (PSE) of the Forsmark area have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The evaluation shows that, even considering remaining uncertainties, the Forsmark area meets all stated safety requirements and preferences. Consequently, from a safety point of view, there is no reason not to continue the Site Investigations of the Forsmark area. There are still uncertainties to resolve and the safety would eventually need to be verified through a full safety assessment. Nevertheless, this Preliminary Safety Evaluation demonstrates that it is likely that a safe repository for spent nuclear fuel of the KBS-3 type could be constructed at the site. The following feedback is provided to the site investigations and the associated site modelling: Reducing the uncertainty on the deformation zone geometry inside the target area would be needed to more firmly define locations of the suitable deposition volumes. There is substantial uncertainty in the Discrete Fracture Network model. Further reduction of the uncertainties, if needed, would probably only be possible from the underground, detailed investigation phase. Efforts need also be spent on improving the DFN-modelling. There are assumptions made in current models that could be challenged and there seems to be room for better use of the borehole information. It is particularly important to

  4. Scrambler therapy efficacy and safety for neuropathic pain correlated with chemotherapy-induced peripheral neuropathy in adolescents: A preliminary study.

    Science.gov (United States)

    Tomasello, Caterina; Pinto, Rita Maria; Mennini, Chiara; Conicella, Elena; Stoppa, Francesca; Raucci, Umberto

    2018-04-06

    Chemotherapy-induced peripheral neuropathy (CIPN) is a common side effect of chemotherapy, in need of effective treatment. Preliminary data support the efficacy of scrambler therapy (ST), a noninvasive cutaneous electrostimulation device, in adults with CIPN. We test the efficacy, safety, and durability of ST for neuropathic pain in adolescents with CIPN. We studied nine pediatric patients with cancer and CIPN who received ST for pain control. Each patient received 45-min daily sessions for 10 consecutive days as a first step, but some of them required additional treatment. Pain significantly improved comparing Numeric Rate Scale after 10 days of ST (9.22 ± 0.83 vs. 2.33 ± 2.34; P < 0.001) and at the end of the optimized cycle (EOC) (9.22 ± 0.83 vs. 0.11 ± 0.33, P < 0.001). The improvement in quality of life was significantly reached on pain interference with general activity (8.67 ± 1.66 vs. 3.33 ± 2.12, P < 0.0001), mood (8.33 ± 3.32 vs. 2.78 ± 2.82, P < 0.0005), walking ability (10.00 vs. 2.78 ± 1.22, P < 0.0001), sleep (7.56 ± 2.24 vs. 2.67 ± 1.41, P < 0.001), and relations with people (7.89 ± 2.03 vs. 2.11 ± 2.03, P < 0.0002; Lansky score 26.7 ± 13.2 vs. 10 days of ST 57.8 ± 13.9, P < 0.001; 26.7 ± 13.2 vs. EOC 71.1 ± 16.2, P < 0.001). Based on these preliminary data, ST could be a good choice for adolescents with CIPN for whom pain control is difficult. ST caused total relief or dramatic reduction in CIPN pain and an improvement in quality of life, durable in follow-up. It caused no detected side effects, and can be retrained successfully. Further larger studies should be performed to confirm our promising preliminary data in pediatric patients with cancer. © 2018 Wiley Periodicals, Inc.

  5. Preliminary analysis of the safety and environmental impact of the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Carlson, R.V.; Jalbert, R.A.

    1980-01-01

    The Tritium Systems Test Assembly (TSTA) is a facility dedicated to the development of technologies associated with the D-T fuel cycle of future fusion reactors while demonstrating that TSTA can be operated safely with no significant losses to the environment. During the initial design stage of TSTA, a safety analysis was performed which investigated the effects of major subsystem component failure, the meteorology and seismicity of the site and their possible effect on the facility, and accident scenarios which result in tritium releases. Major releases of tritium to the environment are considered highly improbable since they require a compound failure of primary and secondary containment, along with either a breach of the building or a failure of the Emergency Tritium Cleanup system. Accidental releases caused by natural phenomena (earthquake, tornado, etc.) are considered highly improbable

  6. Safety and feasibility of inpatient exercise training in pediatric heart failure: a preliminary report.

    Science.gov (United States)

    McBride, Michael G; Binder, Tracy Jo; Paridon, Stephen M

    2007-01-01

    To determine the safety and feasibility of an inpatient exercise training program for a group of pediatric heart transplantation candidates on multiple inotropic support. Children with end-stage heart disease often require heart transplantation. Currently, no data exist on the safety and feasibility of an inpatient exercise training program in pediatric patients awaiting heart transplantation while on inotropic support. Twenty ambulatory patients (11 male; age, 13.6 +/- 3.2 years) were admitted, listed, and subsequently enrolled into an exercise training program while awaiting heart transplantation. Patient diagnoses consisted of dilated cardiomyopathy (n = 15), restrictive cardiomyopathy (n = 1), and failing single-ventricle physiology (n = 4). Inotropic support consisted of a combination of dobutamine, dopamine, or milrinone. Exercise sessions were scheduled three times a week lasting from 30 to 60 minutes and consisted of aerobic and musculoskeletal conditioning. Over 6.2 +/- 4.2 months, 1,251 of a possible 1,508 exercise training sessions were conducted, with a total of 615 hours (26.3 +/- 2.7 min/session) dedicated to low-intensity aerobic exercise. Reasons for noncompliance included a change in medical status, staffing, or patient cooperation. Two adverse episodes (seizures) occurred, neither of which resulted in termination from the program. No adverse episodes of hypotension or significant complex arrhythmias occurred. No complication of medication administration or loss of intravenous access occurred. Data from this study indicate that pediatric patients on inotropic support as a result of systemic ventricular or biventricular heart failure can safely participate in exercise training programs with relatively moderate to high compliance.

  7. Nuclide documentation. Element specific parameter values used in the biospheric models of the safety assessments SR 97 and SAFE

    International Nuclear Information System (INIS)

    Karlsson, Sara; Bergstroem, Ulla

    2002-05-01

    In this report the element and nuclide specific parameter values used in the biospheric models of the safety assessments SR 97 and SAFE are presented. The references used are presented and where necessary the process of estimation of data is described. The parameters treated in this report are distribution coefficients in soil, organic soil and suspended matter in freshwater and brackish water, root uptake factors for pasturage, cereals, root crops and vegetables, bioaccumulation factors for freshwater fish, brackish water fish, freshwater invertebrates and marine water plants, transfer coefficients for transfer to milk and meat, translocation factors and dose coefficients for external exposure, ingestion (age-dependent values) and inhalation (age-dependent values). The radionuclides treated are those which could be of interest in the two safety assessments. Physical data such as half-lives and type of decay are also presented

  8. Nuclide documentation. Element specific parameter values used in the biospheric models of the safety assessments SR 97 and SAFE

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Sara; Bergstroem, Ulla [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    2002-05-01

    In this report the element and nuclide specific parameter values used in the biospheric models of the safety assessments SR 97 and SAFE are presented. The references used are presented and where necessary the process of estimation of data is described. The parameters treated in this report are distribution coefficients in soil, organic soil and suspended matter in freshwater and brackish water, root uptake factors for pasturage, cereals, root crops and vegetables, bioaccumulation factors for freshwater fish, brackish water fish, freshwater invertebrates and marine water plants, transfer coefficients for transfer to milk and meat, translocation factors and dose coefficients for external exposure, ingestion (age-dependent values) and inhalation (age-dependent values). The radionuclides treated are those which could be of interest in the two safety assessments. Physical data such as half-lives and type of decay are also presented.

  9. Ionising radiations. Joint consultative document. Supplementary proposals for provision on radiological protection and draft advice from the National Radiological Protection Board to the Health and Safety Commission

    International Nuclear Information System (INIS)

    1979-01-01

    The consultative document is in two parts. Part 1 indicates the amendments to the first consultative document which would be required in order to implement (in the United Kingdom) the 1978 Draft Euratom Directive (on Basic Safety Standards for the health protection of the general public and workers against the dangers of ionising radiations). Part 2 deals with the system of dose limitation contained within the Euratom Directive. This aspect is discussed, in relation to the Articles of the Directive, under the following headings: limitation of doses for controllable exposures, limits of doses for exposed workers, limitation of doses for apprentices and students, planned special exposures, dose limits for members of the public. The Commission of the European Communities proposals for a draft Directive on Radiological Protection are reproduced as an Appendix, without Annexes. (U.K.)

  10. Safety Assessment Document for the Spent Reactor Fuel Geologic Storage Test in the Climax Granite Stock at the Nevada Test site

    International Nuclear Information System (INIS)

    1980-01-01

    The objective of the Spent Fuel Geologic Storage Test in the Climax Granite Stock is to evaluate the response of a granitic rock mass to the underground storage of encapsulated spent reactor fuel in a geometry that simulates a module of a large-scale geologic repository. This document reports an assessment of the safety of conducting this test. Descriptions are provided of the geography, meteorology, hydrology, geology, and seismology of the Climax Site; the effects of postulated natural phenomena and other activities at the nevada Test Site on the safety of the test; and the design and operation of the test facility and associated equipment. Evaluations are made of both the radiological and nonradiological impacts of normal operations, abnormal operations, and postulated accidents. It is concluded that conduct of the spent fuel test at the Climax Site will not result in any undue risk to the public, property, environment, or site employees

  11. Comparison of AIHA ISO 9001-based occupational health and safety management system guidance document with a manufacturer's occupational health and safety assessment instrument.

    Science.gov (United States)

    Dyjack, D T; Levine, S P; Holtshouser, J L; Schork, M A

    1998-06-01

    Numerous manufacturing and service organizations have integrated or are considering integration of their respective occupational health and safety management and audit systems into the International Organization for Standardization-based (ISO) audit-driven Quality Management Systems (ISO 9000) or Environmental Management Systems (ISO 14000) models. Companies considering one of these options will likely need to identify and evaluate several key factors before embarking on such efforts. The purpose of this article is to identify and address the key factors through a case study approach. Qualitative and quantitative comparisons of the key features of the American Industrial Hygiene Association ISO-9001 harmonized Occupational Health and Safety Management System with The Goodyear Tire & Rubber Co. management and audit system were conducted. The comparisons showed that the two management systems and their respective audit protocols, although structured differently, were not substantially statistically dissimilar in content. The authors recommend that future studies continue to evaluate the advantages and disadvantages of various audit protocols. Ideally, these studies would identify those audit outcome measures that can be reliably correlated with health and safety performance.

  12. Safety and Immunogenicity of an Anti-Zika Virus DNA Vaccine - Preliminary Report.

    Science.gov (United States)

    Tebas, Pablo; Roberts, Christine C; Muthumani, Kar; Reuschel, Emma L; Kudchodkar, Sagar B; Zaidi, Faraz I; White, Scott; Khan, Amir S; Racine, Trina; Choi, Hyeree; Boyer, Jean; Park, Young K; Trottier, Sylvie; Remigio, Celine; Krieger, Diane; Spruill, Susan E; Bagarazzi, Mark; Kobinger, Gary P; Weiner, David B; Maslow, Joel N

    2017-10-04

    Background Although Zika virus (ZIKV) infection is typically self-limiting, other associated complications such as congenital birth defects and the Guillain-Barré syndrome are well described. There are no approved vaccines against ZIKV infection. Methods In this phase 1, open-label clinical trial, we evaluated the safety and immunogenicity of a synthetic, consensus DNA vaccine (GLS-5700) encoding the ZIKV premembrane and envelope proteins in two groups of 20 participants each. The participants received either 1 mg or 2 mg of vaccine intradermally, with each injection followed by electroporation (the use of a pulsed electric field to introduce the DNA sequence into cells) at baseline, 4 weeks, and 12 weeks. Results The median age of the participants was 38 years, and 60% were women; 78% were white, and 22% black; in addition, 30% were Hispanic. At the interim analysis at 14 weeks (i.e., after the third dose of vaccine), no serious adverse events were reported. Local reactions at the vaccination site (e.g., injection-site pain, redness, swelling, and itching) occurred in approximately 50% of the participants. After the third dose of vaccine, binding antibodies (as measured on enzyme-linked immunosorbent assay) were detected in all the participants, with geometric mean titers of 1642 and 2871 in recipients of 1 mg and 2 mg of vaccine, respectively. Neutralizing antibodies developed in 62% of the samples on Vero-cell assay. On neuronal-cell assay, there was 90% inhibition of ZIKV infection in 70% of the serum samples and 50% inhibition in 95% of the samples. The intraperitoneal injection of postvaccination serum protected 103 of 112 IFNAR knockout mice (bred with deletion of genes encoding interferon-α and interferon-β receptors) (92%) that were challenged with a lethal dose of ZIKV-PR209 strain; none of the mice receiving baseline serum survived the challenge. Survival was independent of the neutralization titer. Conclusions In this phase 1, open-label clinical

  13. Posiva's application for a decision in principle concerning a disposal facility for spent nuclear fuel. STUK's statement and preliminary safety appraisal

    Energy Technology Data Exchange (ETDEWEB)

    Ruokola, E. [ed.

    2000-03-01

    In May 1999, Posiva Ltd submitted to the Government an application, pursuant to the Nuclear Energy Act, for a Decision in Principle on a disposal facility for spent nuclear fuel from the Finnish nuclear power plants. The Ministry of Trade and Industry requested the Radiation and Nuclear Safety Authority (STUK) to draw up a preliminary safety appraisal concerning the proposed disposal facility. In the beginning of this report, STUK's statement to the Ministry and Industry concerning the proposed disposal facility is given. In that statement, STUK concludes that the Decision in Principle is currently justified from the standpoint of safety. The statement is followed by a safety appraisal, where STUK deems, how the proposed disposal concept, site and facility comply with the safety requirements included in the Government's Decision (478/1999). STUK's preliminary safety appraisal was supported by contributions from a number of outside experts. A collective opinion by an international group of ten distinguished experts is appended to this report. (orig.)

  14. Preliminary safety analysis report for project 89-GEB-610 Plutonium Finishing Plant instrumentation upgrade. Revision 1

    International Nuclear Information System (INIS)

    Huber, T.E.

    1995-01-01

    This document consists of an analysis of the MICON system upgrade. This project shall install a Micon Co. distributed process monitor and control system with Sparc Sun workstation operator interfaces. The Sparc workstations are housed in consoles custom designed to human factors specifications. The distributed control system (DCS) shall have the installed capacity to monitor and control all related instruments and equipment presently connected to the panels in the PFP Power Control Room 321A as listed in the input/output list. This also includes all devices monitored and controlled by the 2736-ZB Allen Bradley programmable logic controller. The system has since assumed the control and monitoring responsibilities for Projects B- 680H ''Low Level Waste Treatment Facility'' and C-031H ''PFP Liquid Effluent Facilities''. Part of the new en's change area in Building 234-5ZA, Room 712, has been remodeled to house two consoles and one supervisor console. Local control units containing the microprocontrollers and the input/output interface circuit boards shall be wired to the instrumentation and controlled equipment. These units communicate with the Sparc workstations via a redundant data communications highway and shall be strategic, throughout the PFP facility. The DCS has already been purchased from Micon Co., located in Houston Texas, presently on site

  15. Clinical pharmacokinetics, safety, and preliminary efficacy evaluation of icotinib in patients with advanced non-small cell lung cancer.

    Science.gov (United States)

    Liu, Dongyang; Zhang, Li; Wu, Yiwen; Jiang, Ji; Tan, Fenlai; Wang, Yingxiang; Liu, Yong; Hu, Pei

    2015-09-01

    To receive pharmacokinetics, safety, and anti-tumor activity of icotinib, a novel epidermal growth factor receptor (EGFR)-tyrosine kinase inhibitor (TKI), in patients with advanced non-small-cell lung cancer (NSCLC). Patients (n=40) with advanced NSCLC were enrolled to receive escalating doses of icotinib, which was administrated on Day 1 followed by 28-day continuous dosing starting from Day 4. Four dosing regimens, 100mg b.i.d., 150 mg b.i.d., 125 mg t.i.d., and 200mg b.i.d. were studied. Pharmacokinetics (PK), safety, and efficacy of icotinib were evaluated. Icotinib was well tolerated in Chinese patients with refractory NSCLC. No toxicity with >3 grades were reported in more than 2 patients under any dose levels. One complete response (3%) and 9 partial responses (23%) were received. Total disease control rate could reach at 73% and median progress-free survival (range) was 154 (17-462) days. PK exposure of icotinib increased with increase of dose in NSCLC patients. Food was suggested to increase PK exposure by ∼30%. Mean t1/2β was within 5.31-8.07 h. No major metabolite (>10% plasma exposure of icotinib) was found in NSCLC patients. Icotinib with up to 400 mg/day exhibited good tolerance and preliminary antitumor activity in Chinese NSCLC patients. Pharmacokinetics of icotinib and 5 major metabolites were fully investigated in NSCLC patients. Optimized biologic dose (OBD) was finally recommended to be 125 mg t.i.d. for the later clinical study. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  16. ADME studies and preliminary safety pharmacology of LDT5, a lead compound for the treatment of benign prostatic hyperplasia

    Directory of Open Access Journals (Sweden)

    F. Noël

    Full Text Available This study aimed to estimate the absorption, distribution, metabolism and excretion (ADME properties and safety of LDT5, a lead compound for oral treatment of benign prostatic hyperplasia that has previously been characterized as a multi-target antagonist of α1A-, α1D-adrenoceptors and 5-HT1A receptors. The preclinical characterization of this compound comprised the evaluation of its in vitro properties, including plasma, microsomal and hepatocytes stability, cytochrome P450 metabolism and inhibition, plasma protein binding, and permeability using MDCK-MDR1 cells. De-risking and preliminary safety pharmacology assays were performed through screening of 44 off-target receptors and in vivo tests in mice (rota-rod and single dose toxicity. LDT5 is stable in rat and human plasma, human liver microsomes and hepatocytes, but unstable in rat liver microsomes and hepatocytes (half-life of 11 min. LDT5 is highly permeable across the MDCK-MDR1 monolayer (Papp ∼32×10-6 cm/s, indicating good intestinal absorption and putative brain penetration. LDT5 is not extensively protein-bound and is a substrate of human CYP2D6 and CYP2C19 but not of CYP3A4 (half-life >60 min, and did not significantly influence the activities of any of the human cytochrome P450 isoforms screened. LDT5 was considered safe albeit new studies are necessary to rule out putative central adverse effects through D2, 5-HT1A and 5-HT2B receptors, after chronic use. This work highlights the drug-likeness properties of LDT5 and supports its further preclinical development.

  17. Color-coding and human factors engineering to improve patient safety characteristics of paper-based emergency department clinical documentation.

    Science.gov (United States)

    Kobayashi, Leo; Boss, Robert M; Gibbs, Frantz J; Goldlust, Eric; Hennedy, Michelle M; Monti, James E; Siegel, Nathan A

    2011-01-01

    Investigators studied an emergency department (ED) physical chart system and identified inconsistent, small font labeling; a single-color scheme; and an absence of human factors engineering (HFE) cues. A case study and description of the methodology with which surrogate measures of chart-related patient safety were studied and subsequently used to reduce latent hazards are presented. Medical records present a challenge to patient safety in EDs. Application of HFE can improve specific aspects of existing medical chart organization systems as they pertain to patient safety in acute care environments. During 10 random audits over 5 consecutive days (573 data points), 56 (9.8%) chart binders (range 0.0-23%) were found to be either misplaced or improperly positioned relative to other chart binders; 12 (21%) were in the critical care area. HFE principles were applied to develop an experimental chart binder system with alternating color-based chart groupings, simple and prominent identifiers, and embedded visual cues. Post-intervention audits revealed significant reductions in chart binder location problems overall (p < 0.01), for Urgent Care A and B pods (6.4% to 1.2%; p < 0.05), Fast Track C pod (19.3% to 0.0%; p < 0.05) and Behavioral/Substance Abuse D pod (15.7% to 0.0%; p < 0.05) areas of the ED. The critical care room area did not display an improvement (11.4% to 13.2%; p = 0.40). Application of HFE methods may aid the development, assessment, and modification of acute care clinical environments through evidence-based design methodologies and contribute to safe patient care delivery.

  18. The concepts of exclusion, exemption and clearance as used in the interagency basic safety standards and related IAEA documents

    International Nuclear Information System (INIS)

    Webbl, G.A.M.

    1997-01-01

    Mechanisms are needed to remove from regulatory control those exposures or radiation sources that do not warrant concern. In this paper three such conceptual mechanisms are examined from their historical development to their current usage in the Interagency Basic Safety Standards. These concepts are exclusion, applied to exposures that are not amenable to control, exemption applied in advance on the basis of low risks to prevent practices or sources from entering the regulatory control system, and clearance, a similar concept but used to remove sources from the regulatory control system. The application of and interrelationships between these concepts, is described. (author)

  19. [Transjugular intrahepatic portosystemic shunting with covered stents in children: a preliminary study of safety and patency].

    Science.gov (United States)

    Zurera, L J; Espejo, J J; Canis, M; Bueno, A; Vicente, J; Gilbert, J J

    2014-01-01

    To retrospectively analyze the safety and efficacy of transjugular intrahepatic portosystemic shunting (TIPS) using covered stents in children. We present 6 children (mean age, 10.6 years; mean weight, 33.5kg) who underwent TIPS with 8mm diameter Viatorr(®) covered stents for acute (n=4) or recurrent (n=2) upper digestive bleeding that could not be controlled by endoscopic measures. Five of the children had cirrhosis and the other had portal vein thrombosis with cavernous transformation. We analyzed the relapse of upper digestive bleeding, the complications that appeared, and the patency of the TIPS shunt on sequential Doppler ultrasonography or until transplantation. A single stent was implanted in a single session in each child; none of the children died. The mean transhepatic gradient decreased from 16mmHg (range: 12-21mmHg) before the procedure to 9mmHg (range: 1-15mmHg) after TIPS. One patient developed mild encephalopathy, and the girl who had portal vein thrombosis with cavernous transformation developed an acute occlusion of the TIPS that resolved after the implantation of a coaxial stent. Three children received transplants (7, 9, and 10 months after the procedure, respectively), and the patency of the TIPS was confirmed at transplantation. In the three remaining children, patency was confirmed with Doppler ultrasonography 1, 3, and 5 months after implantation. None of the children had new episodes of upper digestive bleeding during follow-up after implantation (mean: 8.1 months). Our results indicate that TIPS with 8mm diameter Viatorr(®) covered stents can be safe and efficacious for the treatment of upper digestive bleeding due to gastroesophageal varices in cirrhotic children; our findings need to be corroborated in larger series. Copyright © 2011 SERAM. Published by Elsevier Espana. All rights reserved.

  20. Study on systematic integration technology of design and safety assessment for HLW geological disposal. 2. Research document

    International Nuclear Information System (INIS)

    Ishihara, Yoshinao; Fukui, Hiroshi; Sagawa, Hiroshi; Matsunaga, Kenichi; Ito, Takaya; Kohanawa, Osamu; Kuwayama, Yuki

    2003-02-01

    The present study was carried out relating to basic design of the Geological Disposal Technology Integration System' that will be systematized as knowledge base for design analysis and safety assessment of HLW geological disposal system by integrating organically and hierarchically various technical information in three study field. The key conclusions are summarized as follows: (1) As referring to the current performance assessment report, the technical information for R and D program of HLW geological disposal system was systematized hierarchically based on summarized information in a suitable form between the work flow (work item) and processes/characteristic flow (process item). (2) As the result of the systematized technical information, database structure and system functions necessary for development and construction to the computer system were clarified in order to secure the relation between technical information and data set for assessment of HLW geological disposal system. (3) The control procedure for execution of various analysis code used by design and safety assessment in HLW geological disposal study was arranged possibility in construction of 'Geological Disposal Technology Integration System' after investigating the distributed computing technology. (author)

  1. Safety

    International Nuclear Information System (INIS)

    1998-01-01

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

  2. Final safety analysis report for the Galileo mission: Volume 3 (Book 2), Nuclear risk analysis document: Appendices: Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-25

    It is the purpose of the NRAD to provide an analysis of the range of potential consequences of accidents which have been identified that are associated with the launching and deployment of the Galileo mission spacecraft. The specific consequences analyzed are those associated with the possible release of radioactive material (fuel) of the Radioisotope Thermoelectric Generators (RTGs). They are in terms of radiation doses to people and areas of deposition of radioactive material. These consequence analyses can be used in several ways. One way is to identify the potential range of consequences which might have to be dealt with if there were to be an accident with a release of fuel, so as to assure that, given such an accident, the health and safety of the public will be reasonably protected. Another use of the information, in conjunction with accident and release probabilities, is to estimate the risks associated with the mission. That is, most space launches occur without incident. Given an accident, the most probable result relative to the RTGs is complete containment of the radioactive material. Only a small fraction of accidents might result in a release of fuel and subsequent radiological consequences. The combination of probability with consequence is risk, which can be compared to other human and societal risks to assure that no undue risks are implied by undertaking the mission. Book 2 contains eight appendices.

  3. Evaluation of geological documents available for provisional safety analyses of potential sites for nuclear waste repositories - Are additional geological investigations needed?

    International Nuclear Information System (INIS)

    2010-10-01

    The procedure for selecting repository sites for all categories of radioactive waste in Switzerland is defined in the conceptual part of the Sectoral Plan for Deep Geological Repositories, which foresees a selection of sites in three stages. In Stage I, Nagra proposed geological siting regions based on criteria relating to safety and engineering feasibility. The Swiss Government (the Federal Council) is expected to decide on the siting proposals in 2011. The objective of Stage 2 is to prepare proposals for the location of the surface facilities within the planning perimeters defined by the Federal Council in its decision on Stage 1 and to identify potential sites. Nagra also has to carry out a provisional safety analysis for each site and a safety-based comparison of the sites. Based on this, and taking into account the results of the socio-economic-ecological impact studies, Nagra then has to propose at least two sites for each repository type to be carried through to Stage 3. The proposed sites will then be investigated in more detail in Stage 3 to ensure that the selection of the sites for the General Licence Applications is well founded. In order to realise the objectives of the upcoming Stage 2, the state of knowledge of the geological conditions at the sites has to be sufficient to perform the provisional safety analyses. Therefore, in preparation for Stage 2, the conceptual part of the Sectoral Plan requires Nagra to clarify the need for additional investigations aimed at providing input for the provisional safety analyses. The purpose of the present report is to document Nagra's technical-scientific assessment of this need. The focus is on evaluating the geological information based on processes and parameters that are relevant for safety and engineering feasibility. In evaluating the state of knowledge the key question is whether additional information could lead to a different decision regarding the selection of the sites to be carried through to Stage 3

  4. WNA's Policy Document : sustaining global best practices in uranium, mining and processing, principles for managing radiation, health and safety, waste and the environment

    International Nuclear Information System (INIS)

    Saint-Pierre, S.; Waste Management and Decommissioning Working Group-WM and DW

    2008-01-01

    The worldwide community of uranium mining and processing recognizes that managing radiation, health and safety, waste and the environment is paramount. Such responsible management applies at all stages of planning and activities. Today we are acting to ensure that all parties directly involved in uranium mining and processing strive to achieve the highest levels of excellence in these fields. We are doing so by sustaining a strong safety culture based on a commitment to common, internationally shared principles. This paper sets out principles for the management of radiation, health and safety, waste and the environment applicable to sites throughout the world. In national and regional settings where nuclear fuel cycle activities are well developed, these principles already serve as the underpinning for 'Codes of Practice' that govern uranium mining and processing. In any given setting, a Code of Practice is needed to guide practical implementation of these principles according to the regional, national or site-specific context. These principles are published in the belief that they hold special relevance for emerging uranium producing countries that do not yet have fully developed regulations for the control of radiation, health and safety, waste and the environment associated with uranium mining and processing. The principles are equally relevant for operators, contractors, and regulators newly engaged in uranium mining and processing. Once national regulations are fully developed, they can be expected to embody these principles. Each principle affirmed here will not apply to the same extent for each party. Ultimately, the precise allocation of responsibilities must be set at the national and local levels. This document holds the status of a policy and ethical declaration by the full WNA membership, which the global nuclear industry. The principles affirmed here are supported by key relevant international organizations, including the IAEA and the global mining

  5. Biosphere modelling for safety assessment of geological disposal taking account of denudation of contaminated soils. Research document

    International Nuclear Information System (INIS)

    Kato, Tomoko

    2003-03-01

    Biosphere models for safety assessment of geological disposal have been developed on the assumption that the repository-derived radionuclides reach surface environment by groundwater. In the modelling, river, deep well and marine have been considered as geosphere-biosphere (GBIs) and some Japanese-specific ''reference biospheres'' have been developed using an approach consistent with the BIOMOVS II/BIOMASS Reference Biosphere Methodology. In this study, it is assumed that the repository-derived radionuclide would reach surface environment in the form of solid phase by uplift and erosion of contaminated soil and sediment. The radionuclides entered into the surface environment by these processes could be distributed between solid and liquid phases and could spread within the biosphere via solid phase and also liquid phase. Based on these concepts, biosphere model that considers variably saturated zone under surface soil (VSZ) as a GBI was developed for calculating the flux-to-dose conversion factors of three exposure groups (farming, freshwater fishing, marine fishing) based on the Reference Biosphere Methodology. The flux-to-dose conversion factors for faming exposure group were the highest, and ''inhalation of dust'', external irradiation from soil'' and ''ingestion of soil'' were the dominant exposure pathways for most of radionuclides considered in this model. It is impossible to compare the flux-to-dose conversion factors calculated by the biosphere model in this study with those calculated by the biosphere models developed in the previous studies because the migration processes considered when the radionuclides entered the surface environment through the aquifer are different among the models; i.e. it has been assumed that the repository-derived radionuclides entered the GBIs such as river, deep well and marine via groundwater without dilution and retardation at the aquifer in the previous biosphere models. Consequently, it must be modelled the migration of

  6. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  7. RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

    International Nuclear Information System (INIS)

    KOZLOWSKI, S.D.

    2007-01-01

    This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditions for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below

  8. Preliminary I&C Design for LORELEI

    International Nuclear Information System (INIS)

    Korotkin, S.; Kaufman, Y.; Guttmann, E. B.; Levy, S.; Amidan, D.; Gdalyho, B.; Cahana, T.; Ellenbogen, A.; Arad, M.; Weiss, Y.; Sasson, A.; Ferry, L.; Bourrelly, F.; Cohen, Y.

    2014-01-01

    This document summarizes the preliminary I&C design for LORELEI experiment The preliminary design deals with considerations regarding appropriate safety and service instrumentation. The determined closed loop control rules for temperature and position will be implemented in the detailed design. The Computer Aided Operator Decisions System (CAODS) will be used for prediction of hot spot temperature and thickness of oxidation layer using Baker-Just correlation. The proposed hybrid simulation system comprising of both virtual and real hardware will be in-cooperated for LORELEI verification. It will perform both integration cold tests for a partial hardware loop and virtual tests for the final I&C design

  9. Nuclear criticality safety guide

    International Nuclear Information System (INIS)

    Pruvost, N.L.; Paxton, H.C.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators

  10. Nuclear criticality safety guide

    Energy Technology Data Exchange (ETDEWEB)

    Pruvost, N.L.; Paxton, H.C. [eds.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

  11. Preliminary report of radiological safety to hydrology 1993 campaign; Informe preliminar de seguridad radiologica para la campana de hidrologia 1993

    Energy Technology Data Exchange (ETDEWEB)

    Badano, A; Suarez Antola, R; Dellepere, A; Barreiro, M [MIEM, Direccion Nacional de Tecnologia Nuclear, Montevideo (Uruguay)

    1993-07-01

    This report has been prepared based on the interaction between project managers and division radiological Protection and Nuclear Safety. In seeking to establish a basis for approval from the point of view of radiation safety practices . The idea for the audit has been provided at all times because the interest was the exchange of ideas and the use of common sense to improve the safety of radioactive substances, security of operators and public safety and environment.The above shows that in the planned radiation safety condition described in this report,the practice can be carried out according to the criteria of safety accepted .

  12. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

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

  13. Joint statement by the chairmen of the Standing Committee on Reactors and the Reactor Safety Commission on safety-related documents drawn up under the Franco-German Commission on Safety Questions for Nuclear Installations (DFK). August 29, 1986

    International Nuclear Information System (INIS)

    1986-01-01

    The report contains: 1. The present situation regarding hints and recommendations in the safety declaration by the TUEV Baden on the subject of Cattenom nuclear power station in June 1982. 2. The present situation regarding hints and recommendations in the declaration by the TUEV Rhineland in February 1982 on the subject of radiological aspects and features of the Cattenom site for the nuclear power plant. 3. Answers to press allegations about the safety of Cattenom NPP: scram system, safety cooling system, heat sink for after-heat removal, emergency power supplies, load-follow operation, air crashes, susceptibility to failures. (orig./HP) [de

  14. Immunogenicity and Safety of the New Inactivated Quadrivalent Influenza Vaccine Vaxigrip Tetra: Preliminary Results in Children ≥6 Months and Older Adults

    Directory of Open Access Journals (Sweden)

    Emanuele Montomoli

    2018-03-01

    Full Text Available Since the mid-1980s, two lineages of influenza B viruses have been distinguished. These can co-circulate, limiting the protection provided by inactivated trivalent influenza vaccines (TIVs. This has prompted efforts to formulate quadrivalent influenza vaccines (QIVs, to enhance protection against circulating influenza B viruses. This review describes the results obtained from seven phase III clinical trials evaluating the immunogenicity, safety, and lot-to-lot consistency of a new quadrivalent split-virion influenza vaccine (Vaxigrip Tetra® formulated by adding a second B strain to the already licensed TIV. Since Vaxigrip Tetra was developed by means of a manufacturing process strictly related to that used for TIV, the data on the safety profile of TIV are considered supportive of that of Vaxigrip Tetra. The safety and immunogenicity of Vaxigrip Tetra were similar to those of the corresponding licensed TIV. Moreover, the new vaccine elicits a superior immune response towards the additional strain, without affecting immunogenicity towards the other three strains. Vaxigrip Tetra is well tolerated, has aroused no safety concerns, and is recommended for the active immunization of individuals aged ≥6 months. In addition, preliminary data confirm its immunogenicity and safety even in children aged 6–35 months and its immunogenicity in older subjects (aged 66–80 years.

  15. Geoscientific long-term prognosis. Preliminary safety analysis for the site Gorleben; Geowissenschaftliche Langzeitprognose. Bericht zum Arbeitspaket 2. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

    Energy Technology Data Exchange (ETDEWEB)

    Mrugalla, Sabine [Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover (Germany)

    2011-07-15

    The preliminary safety analysis of the site Gorleben includes the following chapters: (1) Introduction; (2) Aim and content of the geoscientific long-term prognosis for the site Gorleben; (3) Boundary conditions at the site Gorleben: climate; geomorphology; overlying rocks and adjoining rocks; hydrogeology; salt deposit Gorleben. (4) Probable future geological developments at the site Gorleben: supraregional developments with effects on the site Gorleben; glacial period developments; developments of the geomorphology, overlying and adjoining rocks; future developments of the hydrological systems at the site Gorleben; future saliniferous specific developments of the salt deposit Gorleben. (5) Commentary on the unlikely or excludable developments of the site Gorleben.

  16. When doctors share visit notes with patients: a study of patient and doctor perceptions of documentation errors, safety opportunities and the patient-doctor relationship.

    Science.gov (United States)

    Bell, Sigall K; Mejilla, Roanne; Anselmo, Melissa; Darer, Jonathan D; Elmore, Joann G; Leveille, Suzanne; Ngo, Long; Ralston, James D; Delbanco, Tom; Walker, Jan

    2017-04-01

    Patient advocates and safety experts encourage adoption of transparent health records, but sceptics worry that shared notes may offend patients, erode trust or promote defensive medicine. As electronic health records disseminate, such disparate views fuel policy debates about risks and benefits of sharing visit notes with patients through portals. Presurveys and postsurveys from 99 volunteer doctors at three US sites who participated in OpenNotes and postsurveys from 4592 patients who read at least one note and submitted a survey. Patients read notes to be better informed and because they were curious; about a third read them to check accuracy. In total, 7% (331) of patients reported contacting their doctor's office about their note. Of these, 29% perceived an error, and 85% were satisfied with its resolution. Nearly all patients reported feeling better (37%) or the same (62%) about their doctor. Patients who were older (>63), male, non-white, had fair/poor self-reported health or had less formal education were more likely to report feeling better about their doctor. Among doctors, 26% anticipated documentation errors, and 44% thought patients would disagree with notes. After a year, 53% believed patient satisfaction increased, and 51% thought patients trusted them more. None reported ordering more tests or referrals. Despite concerns about errors, offending language or defensive practice, transparent notes overall did not harm the patient-doctor relationship. Rather, doctors and patients perceived relational benefits. Traditionally more vulnerable populations-non-white, those with poorer self-reported health and those with fewer years of formal education-may be particularly likely to feel better about their doctor after reading their notes. Further informing debate about OpenNotes, the findings suggest transparent records may improve patient satisfaction, trust and safety. Published by the BMJ Publishing Group Limited. For permission to use (where not already

  17. ENSI's view on NTB-10-01 'Evaluation of the geological documents for the provisional safety assessment in SGT Stage 2' - Sectoral Plan 'Deep Geological Disposal'

    International Nuclear Information System (INIS)

    2011-03-01

    As a preliminary action ahead of Stage 2 in the Sectoral Plan 'Deep Geological Disposal' ('Sachplan Geologische Tiefenlager': SGT) the bodies in charge of the management of the radioactive wastes have to determine, together with the Swiss Federal Nuclear Safety Inspectorate (ENSI), which complementary investigations are needed for the provisional safety assessment of the foreseen repositories. In particular, comparisons between the different sites must be possible. If some doubts remain, new investigations have to be conducted. In SGT Stage 1 the National Cooperative for the Disposal of Radioactive Waste (NAGRA) stated that a time period of 100'000 years shall be considered for the storage of low- and intermediate-level radioactive wastes (SMA) and of 1 million years for the storage of high-level wastes (HAA). In SGT Stage 2, considerations about possible modifications of the biosphere during the considered time periods for SMA as well as for HAA have to be included. In what regards the host rock, ENSI declares that the dataset describing the Opalinus clay is sufficient for the site-specific safety analyses. What regards the Wellenberg site for a SMA repository, ENSI has earlier analysed the license request of the NAGRA. For the Mergel formations of the Helveticum, NAGRA conducted a complete safety analysis. ENSI considers the knowledge acquired for these rock types as sufficient for the technical safety comparisons in SGT Stage 2. The knowledge of the rock 'Brauner Dogger' is correct but some more information about the lithostratigraphy and the biostratigraphy is requested. The understanding of the effect of the rock density reduction on the hydraulic conductibility of Opalinus clay and Mergel formations is accurate enough for the comparisons in SGT Stage 2. For the 'Brauner Dogger', comparable data are missing, but the proposed thickness of 300 m of the host rock offers a sufficient protection against density reduction effects. Concerning the behaviour of the

  18. Analysis of the prevailing standard and technical documents for their applicability under conditions of the 'Ukrytie' operation and technical proposals relating to the elaboration of new ones concerning fire safety

    International Nuclear Information System (INIS)

    Nazarenko, B.S.; Emets, V.G.

    1998-01-01

    An analysis of the prevailing laws and standards and technical documents (STD) to ensure safe operation of nuclear power installations, with requirements of nuclear, radiation and fire safety taken into account, has been performed. Proposals on application of some items of prevailing STD under conditions of the 'Ukrytie' operation are presented. Also given are technical proposals on correction of the prevailing operational standard documents and elaboration of special STD

  19. Dataset for Phase I randomized clinical trial for safety and tolerability of GET 73 in single and repeated ascending doses including preliminary pharmacokinetic parameters

    Directory of Open Access Journals (Sweden)

    Carolina L. Haass-Koffler

    2017-12-01

    Full Text Available The data in this article outline the methods used for the administration of GET 73 in the first time-in-human manuscript entitled “Phase I randomized clinical trial for the safety, tolerability and preliminary pharmacokinetics of the mGluR5 negative allosteric modulator GET 73 following single and repeated doses in healthy male volunteers” (Haass-Koffler et al., 2017 [1]. Data sets are provided in two different manners. The first series of tables provided includes procedural information about the experiments conducted. The next series of tables provided includes Pharmacokinetic (PK parameters for GET 73 and its main metabolite MET 2. This set of data is comprised by two experiments: Experiment 1 references a single ascending dose administration of GET 73 and Experiment 2 references a repeated ascending dose administration of GET 73. Keywords: Glutamate receptor subtype 5 (mGlu5, Allosteric modulator, GET 73, Safety, Tolerability

  20. Preliminary study to questions relating to the safety of nuclear power plants A and B at Biblis

    International Nuclear Information System (INIS)

    Fischer, B.; Hahn, L.; Rausch, L.

    1985-01-01

    With a view to developing suitable tools for the safety evaluation of reactors A and B at Biblis, the publication compiles all aspects relevant to safety, creates an evaluation frame, and evaluates the aspects relevant to safety by means of this frame of evaluation. According to the composition of the work, the overall subject is split up into the complexes information, acquisition, evaluation of operational experience, probabilistic analyses, comparison with newer PWR type reactors, fulfilling of injunctions, modifications due to disposal problems, and the disposal situation. (DG) [de

  1. Applications for electronic documents

    International Nuclear Information System (INIS)

    Beitel, G.A.

    1995-01-01

    This paper discusses the application of electronic media to documents, specifically Safety Analysis Reports (SARs), prepared for Environmental Restoration and Waste Management (ER ampersand WM) programs being conducted for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). Efforts are underway to upgrade our document system using electronic format. To satisfy external requirements (DOE, State, and Federal), ER ampersand WM programs generate a complement of internal requirements documents including a SAR and Technical Safety Requirements along with procedures and training materials. Of interest, is the volume of information and the difficulty in handling it. A recently prepared ER ampersand WM SAR consists of 1,000 pages of text and graphics; supporting references add 10,000 pages. Other programmatic requirements documents consist of an estimated 5,000 pages plus references

  2. Importance of the documentation of the manual of quality and procedures handbook in the nuclear technology center

    International Nuclear Information System (INIS)

    Domech More, J.; Bolanos Hernandez, R.; Quitero Rosello, R.; Fernandez Rondon, M.; Milian Lorenzo, D.; Rodriguez Gual, M.

    1997-01-01

    In the present work is presented the methodology used for the elaboration of manual of quality of the Nuclear Technology Center and the technical Procedures Handbook for the execution of Preliminary Safety report of the Juragua Nuclear Power Plant, as well as the importance that has this documentation for the work of the center

  3. A Preliminary Study on the Measures to Assess the Organizational Safety: The Cultural Impact on Human Error Potential

    International Nuclear Information System (INIS)

    Lee, Yong Hee; Lee, Yong Hee

    2011-01-01

    The Fukushima I nuclear accident following the Tohoku earthquake and tsunami on 11 March 2011 occurred after twelve years had passed since the JCO accident which was caused as a result of an error made by JCO employees. These accidents, along with the Chernobyl accident, associated with characteristic problems of various organizations caused severe social and economic disruptions and have had significant environmental and health impact. The cultural problems with human errors occur for various reasons, and different actions are needed to prevent different errors. Unfortunately, much of the research on organization and human error has shown widely various or different results which call for different approaches. In other words, we have to find more practical solutions from various researches for nuclear safety and lead a systematic approach to organizational deficiency causing human error. This paper reviews Hofstede's criteria, IAEA safety culture, safety areas of periodic safety review (PSR), teamwork and performance, and an evaluation of HANARO safety culture to verify the measures used to assess the organizational safety

  4. A Preliminary Study on the Measures to Assess the Organizational Safety: The Cultural Impact on Human Error Potential

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Hee; Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The Fukushima I nuclear accident following the Tohoku earthquake and tsunami on 11 March 2011 occurred after twelve years had passed since the JCO accident which was caused as a result of an error made by JCO employees. These accidents, along with the Chernobyl accident, associated with characteristic problems of various organizations caused severe social and economic disruptions and have had significant environmental and health impact. The cultural problems with human errors occur for various reasons, and different actions are needed to prevent different errors. Unfortunately, much of the research on organization and human error has shown widely various or different results which call for different approaches. In other words, we have to find more practical solutions from various researches for nuclear safety and lead a systematic approach to organizational deficiency causing human error. This paper reviews Hofstede's criteria, IAEA safety culture, safety areas of periodic safety review (PSR), teamwork and performance, and an evaluation of HANARO safety culture to verify the measures used to assess the organizational safety

  5. Preliminary statement concerning the expert opinion and other documents relating to the Wackersdorf fuel reprocessing plant, laid open between June 18 and July 15, 1985 by the Bavarian Ministry of State for Regional Development and Environmental Affairs

    International Nuclear Information System (INIS)

    1985-01-01

    The conclusion drawn in the statement is that the expert opinions and documents available do not justify the granting of a first partial licence for the erection of the fuel reprocessing plant. Such a first partial licence would include the site approval, the outline design approval, and part-construction permits, in this case for instance for the fuel element receiving store. The site approval would be a premature decision according to this statement, as some decisive expert opinions (on meteorological, hydrological, and radioecological site conditions) are not yet completed. The documents available for evaluation of the design concept are said to have been insufficient in many regards, so that all the authors can state at the present time is that 'there may be technically feasible solutions by now'. This is judged to be an insufficient basis for an outline design approval in accordance with the Atomic Energy Act. The reasons brought forward against licensing the fuel element receiving store are similar. For lack of appropriate documents, the authors of the statement found themselves unable to assess the safety of a major component, the fuel element storage tank. (orig./HSCH) [de

  6. A guide to introducing burnup credit, preliminary version (English translation)

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

    2017-06-01

    There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)

  7. Preliminary Study for Application of the New Safety Goal related with the Limitation of Cs-137 release

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro; Shin, Tae Young [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    In the New Safety Goal, it is clearly stated that the Probabilistic Safety Assessment (PSA) should be performed with the proper technical appropriateness, the detail, and the scope in accordance with the integrated risk assessment against the accident for the nuclear power plants. This requirement is known to be come from the provision for preventing the long term ground contamination due to the release of radioactive material. However, there were so many concerns that this goal is so severe that the current design, even in the case of the constructing nuclear power plants, cannot meet this criterion. Especially for the operating nuclear power plants, since there were no mitigation facilities against the severe accident at the design stages, the application of this new goal is known to be much severe that the constructing nuclear power plants and it is necessary to develop the alternative methods to strengthen the safety of the operating nuclear power plants. The purpose of this study is to review the new safety goal from the view point of severe accident analysis and probabilistic safety assessment, and to find the appropriate methods in order to meet that goal for the operating nuclear power plants. In order to strengthen the safety for domestic nuclear power plants, all of the domestic operating nuclear power plants are required to prepare the Accident Management Plan within 3 years. Also, this Accident Management Plan should meet the New Safety Goal including the requirement that the sum of the accident frequency that the release of the radioactive nuclide Cs- 137 to the environment exceeds the 100TBq should be less than 1.0E-6/RY. Since the operating nuclear power plants was not designed against the severe accident and they have the limited exclusive mitigation facilities, it is not easy to meet the New Safety Goal. So, it is necessary to develop the alternative methods to meet the New Safety Goal. In this study, the amount of Cs-137 released to the

  8. Preliminary Study for Application of the New Safety Goal related with the Limitation of Cs-137 release

    International Nuclear Information System (INIS)

    Seo, Mi Ro; Shin, Tae Young

    2016-01-01

    In the New Safety Goal, it is clearly stated that the Probabilistic Safety Assessment (PSA) should be performed with the proper technical appropriateness, the detail, and the scope in accordance with the integrated risk assessment against the accident for the nuclear power plants. This requirement is known to be come from the provision for preventing the long term ground contamination due to the release of radioactive material. However, there were so many concerns that this goal is so severe that the current design, even in the case of the constructing nuclear power plants, cannot meet this criterion. Especially for the operating nuclear power plants, since there were no mitigation facilities against the severe accident at the design stages, the application of this new goal is known to be much severe that the constructing nuclear power plants and it is necessary to develop the alternative methods to strengthen the safety of the operating nuclear power plants. The purpose of this study is to review the new safety goal from the view point of severe accident analysis and probabilistic safety assessment, and to find the appropriate methods in order to meet that goal for the operating nuclear power plants. In order to strengthen the safety for domestic nuclear power plants, all of the domestic operating nuclear power plants are required to prepare the Accident Management Plan within 3 years. Also, this Accident Management Plan should meet the New Safety Goal including the requirement that the sum of the accident frequency that the release of the radioactive nuclide Cs- 137 to the environment exceeds the 100TBq should be less than 1.0E-6/RY. Since the operating nuclear power plants was not designed against the severe accident and they have the limited exclusive mitigation facilities, it is not easy to meet the New Safety Goal. So, it is necessary to develop the alternative methods to meet the New Safety Goal. In this study, the amount of Cs-137 released to the

  9. Documentation: Records and Reports.

    Science.gov (United States)

    Akers, Michael J

    2017-01-01

    This article deals with documentation to include the beginning of documentation, the requirements of Good Manufacturing Practice reports and records, and the steps that can be taken to minimize Good Manufacturing Practice documentation problems. It is important to remember that documentation for 503a compounding involves the Formulation Record, Compounding Record, Standard Operating Procedures, Safety Data Sheets, etc. For 503b outsourcing facilities, compliance with Current Good Manufacturing Practices is required, so this article is applicable to them. For 503a pharmacies, one can see the development and modification of Good Manufacturing Practice and even observe changes as they are occurring in 503a documentation requirements and anticipate that changes will probably continue to occur. Copyright© by International Journal of Pharmaceutical Compounding, Inc.

  10. Standardization Documents

    Science.gov (United States)

    2011-08-01

    Specifications and Standards; Guide Specifications; CIDs; and NGSs . Learn. Perform. Succeed. STANDARDIZATION DOCUMENTS Federal Specifications Commercial...national or international standardization document developed by a private sector association, organization, or technical society that plans ...Maintain lessons learned • Examples: Guidance for application of a technology; Lists of options Learn. Perform. Succeed. DEFENSE HANDBOOK

  11. Radiological consequence analysis. Report on the work package 10. Preliminary safety analysis for the site Gorleben; Radiologische Konsequenzanalyse. Bericht zum Arbeitspaket 10. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

    Energy Technology Data Exchange (ETDEWEB)

    Larue, Juergen; Baltes, Bruno; Fischer, Heidemarie; Frieling, Gerd; Kock, Ingo; Navarro, Martin; Seher, Holger

    2013-02-15

    Work package 10 of the preliminary safety analysis for the site Gorleben concerns the analysis of release scenarios and the radiological consequences. The report includes the description of the applied methodology, the concept for the mathematical description of scenarios, the variety of scenarios for the radiological consequence analysis with the definition of the reference scenario. The database for the calculation covers geology and hydrology, the final repository concept and the process of the final disposal system. The used models were the one-phase transport model MARINE and the two-phase transport model TOUGH2. The report summarizes the results including an evaluation of the realization of the demonstration concept with respect to solution transport, fluid transport, robustness of the modeling results and analogy discussion to other disposal variants (i.e. borehole disposal and storage of transport or storage casks).

  12. Dataset for Phase I randomized clinical trial for safety and tolerability of GET 73 in single and repeated ascending doses including preliminary pharmacokinetic parameters.

    Science.gov (United States)

    Haass-Koffler, Carolina L; Goodyear, Kimberly; Long, Victoria M; Tran, Harrison H; Loche, Antonella; Cacciaglia, Roberto; Swift, Robert M; Leggio, Lorenzo

    2017-12-01

    The data in this article outline the methods used for the administration of GET 73 in the first time-in-human manuscript entitled "Phase I randomized clinical trial for the safety, tolerability and preliminary pharmacokinetics of the mGluR5 negative allosteric modulator GET 73 following single and repeated doses in healthy male volunteers" (Haass-Koffler et al., 2017) [1]. Data sets are provided in two different manners. The first series of tables provided includes procedural information about the experiments conducted. The next series of tables provided includes Pharmacokinetic (PK) parameters for GET 73 and its main metabolite MET 2. This set of data is comprised by two experiments: Experiment 1 references a single ascending dose administration of GET 73 and Experiment 2 references a repeated ascending dose administration of GET 73.

  13. Lifecycle management for nuclear engineering project documents

    International Nuclear Information System (INIS)

    Zhang Li; Zhang Ming; Zhang Ling

    2010-01-01

    The nuclear engineering project documents with great quantity and various types of data, in which the relationships of each document are complex, the edition of document update frequently, are managed difficultly. While the safety of project even the nuclear safety is threatened seriously by the false documents and mistakes. In order to ensure the integrality, veracity and validity of project documents, the lifecycle theory of document is applied to build documents center, record center, structure and database of document lifecycle management system. And the lifecycle management is used to the documents of nuclear engineering projects from the production to pigeonhole, to satisfy the quality requirement of nuclear engineering projects. (authors)

  14. Development of a Preliminary Model for Evaluating Occupational Health and Safety Risk Management Maturity in Small and Medium-Sized Enterprises

    Directory of Open Access Journals (Sweden)

    Bilal Kaassis

    2018-02-01

    Full Text Available Management of occupational health and safety (OHS risks is a crucial component of any business. Numerous investigations have shown that work-related injuries and deaths occur disproportionately in small-to-medium-sized enterprises (SMEs and that this is clearly due to deficient management of OHS risks. The main goal of this work is to develop a base of indicators suitable for evaluating OHS risk management maturity in industrial SMEs. A preliminary model is then proposed for this evaluation, based on a small number of relevant indicators selected from a careful bibliographic review. The work begins with a critical review of the literature and analysis of known concepts, methods, tools and models of measurement of risk analysis maturity in order to extract relevant indicators. The most suitable indicators are then grouped to form the basis of a preliminary model for evaluating OHS risk management maturity in the SME setting. Our findings will help managers of SMEs make sound decisions in their quest to improve the OHS performance of their businesses.

  15. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

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

  16. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

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

  17. Preliminary safety assessments in construction of the pilot industrial facility for final disposal of low and intermediate radioactive waste in the archipelago Novaya Zemlya

    International Nuclear Information System (INIS)

    Lopatin, V.V.; Lobanov, N.F.; Mankin, V.I.; Karamushka, V.P.; Ostroborodov, V.V.

    1999-01-01

    This presentation discusses a preliminary safety evaluation of radioactive waste burial at the experimental plant located on Novaya Zemlya. The issues considered are (1) the main provisions on radioactive waste burial in permafrost rock, (2) mining, geological and geocryological conditions at the experimental works' operating site, (3) the main properties of solid and solidified radioactive wastes, (4) the main parameters of the experimental works, (5) preliminary evaluation of safety. The evaluation includes the main requirements to geocryologic characteristics of the permafrost rock intended for waste burial and analyses the seasonal mining-geological and geocryological conditions in the area of the experimental works. The area is situated within the limits of the southern Novozemelsky anticlinorium composed of the Silurian, Devonian and carboniferous rocks of the Paleozoic group. It is mainly limestone and dolomite, showing in rock sequence the layers, benches and horizons of clay shales, aleurolites, conglomerates and magmatic rocks covered with a thin Quaternary sedimentary mantle on the surface. The area is characterised by a confluent continuous layer no less than 300 m thick, seasonal thawing depth 0.5-2.0 m, annual zero temperature variations 10-15 m by the depth, and mean annual rock temperature of -4.5 - 5.0 C. The plant is an independent enterprise supplied with all the required services for industrial and communal/living purposes. The evaluation studies two possible scenarios for accidents during transport of waste to Novaya Zemlya, and the consequences of damage to the plant caused by the impact of a celestial body/flying object, by a catastrophic earthquake, and the effect of global climate warming in the Arctic area

  18. Study of Occupational Safety and Health Audit on Facilities at Ungku Omar College, Universiti Kebangsaan Malaysia (UKM): A Preliminary Analysis

    Science.gov (United States)

    Ariffin, Kadir; Ahmad, Shaharuddin; Aiyub, Kadaruddin; Awang, Azhan; Aziz, Azmi; Mohamad, Lukman Z.; Mamat, Samsu Adabi

    2010-01-01

    Occupational safety and health (OSH) in Universiti Kebangsaan Malaysia (UKM) is being considered as an important program to measure employee and student welfare and well-being. During academic session, apart from attending lectures, laboratory works, tutorial and library search, majority of students spend most of their time in residential…

  19. A two-step approach for the preliminary evaluation of the thermal-hydraulics and safety of the ELSY open square core design

    International Nuclear Information System (INIS)

    Meloni, Paride; Bandini, Giacomino; Polidori, Massimiliano; Cervone, Antonio; Manservisi, Sandro

    2009-01-01

    Several innovative solutions for a liquid metal fast reactor design have been investigated in the EURATOM Sixth Framework Programme and an open-assembly core design for the ELSY (European Lead-cooled System) reactor has been proposed by ENEA. The development of this new reactor, based on innovative neutronic and safety considerations, requires a new approach to the thermal-hydraulic (T/H) core design. In this paper a new two-step approach of the T/H analysis for this open-assembly core is presented and, in particular is used for the evaluation of the preliminary core design of a 1500 MW lead fast reactor with open square lattice and three fuel radial zones with different levels of enrichment. In the first step a preliminary thermal-hydraulic and safety evaluation of the core neutronic design is investigated by using a one-dimensional RELAP5 model for independent channel analysis. Then two and three-dimensional effects are taken into account by using a dedicated tool for the evaluation of assembly mixing effects. The RELAP5 model, based on pressure loss and heat transfer correlations available for heavy liquid metal flows in rod bundle, consists of completely independent assemblies and therefore it can be used for a conservative evaluation of the thermal-hydraulics of the core reactor. Due to the open-lattice configuration, the two and three-dimensional effects are important and they are taken into account by using a simplified three-dimensional numerical model of an open square lattice reactor core, developed with the purpose of analyzing the whole core behavior. The numerical simulation is performed at assembly length level taking into account the local fluctuations of turbulent viscosity and energy exchange coefficients at sub-channel level through transfer operators based on parametric coefficients. A preliminary evaluation of the mixing effects between assembly flows on the temperature field has been performed by using an average assembly turbulent viscosity

  20. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

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

    2000-05-01

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

  1. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  2. Maury Documentation

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Supporting documentation for the Maury Collection of marine observations. Includes explanations from Maury himself, as well as guides and descriptions by the U.S....

  3. Documentation Service

    International Nuclear Information System (INIS)

    Charnay, J.; Chosson, L.; Croize, M.; Ducloux, A.; Flores, S.; Jarroux, D.; Melka, J.; Morgue, D.; Mottin, C.

    1998-01-01

    This service assures the treatment and diffusion of the scientific information and the management of the scientific production of the institute as well as the secretariat operation for the groups and services of the institute. The report on documentation-library section mentions: the management of the documentation funds, search in international databases (INIS, Current Contents, Inspects), Pret-Inter service which allows accessing documents through DEMOCRITE network of IN2P3. As realizations also mentioned are: the setup of a video, photo database, the Web home page of the institute's library, follow-up of digitizing the document funds by integrating the CD-ROMs and diskettes, electronic archiving of the scientific production, etc

  4. CERN's new safety policy

    CERN Multimedia

    2014-01-01

    The documents below, published on 29 September 2014 on the HSE website, together replace the document SAPOCO 42 as well as Safety Codes A1, A5, A9, A10, which are no longer in force. As from the publication date of these documents any reference made to the document SAPOCO 42 or to Safety Codes A1, A5, A9 and A10 in contractual documents or CERN rules and regulations shall be deemed to constitute a reference to the corresponding provisions of the documents listed below.   "The CERN Safety Policy" "Safety Regulation SR-SO - Responsibilities and organisational structure in matters of Safety at CERN" "General Safety Instruction GSI-SO-1 - Departmental Safety Officer (DSO)" "General Safety Instruction GSI-SO-2 - Territorial Safety Officer (TSO)" "General Safety Instruction GSI-SO-3 - Safety Linkperson (SLP)" "General Safety Instruction GSI-SO-4 - Large Experiment Group Leader In Matters of Safety (LEXGLI...

  5. Interim safety basis compliance matrix for Trenches 31 and 34

    International Nuclear Information System (INIS)

    Ames, R.R.

    1994-01-01

    The tables provided in this document identify the specific requirements and basis for the administrative controls established in the Westinghouse Hanford Company (WHC) Solid Waste Burial Ground (SWBG) Interim Safety Basis (ISB) for operation of the Project W-025, Mixed Waste Lined Landfill (Trenches 31 and 34). The tables document the necessary controls and implementing procedures to ensure compliance with the requirements of the ISB. These requirements provide a basis for future Unreviewed Safety Questions (USQ) screening of applicable procedure changes, proposed physical modifications, tests, experiments, and occurrences. Table 1 provides the SWBG interim Operational Safety Requirements administrative controls matrix. The specific assumptions and commitments used in the safety analysis documents applicable to disposal of mixed wastes in Trenches 31 and 34 are provided in Table 2. Table 3 is provided to document the potential engineered and administrative mitigating features identified in the Preliminary Hazard Analysis (PHA) for disposal of mixed waste

  6. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  7. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  8. Preliminary Study on Effect of Aviation Fuel in the Safety Evaluation of Nuclear Power Plant Crashed by Aircraft

    International Nuclear Information System (INIS)

    Jin, Byeong Moo; Jeon, Se Jin; Lee, Yun Seok; Kim, Young Jin

    2011-01-01

    As the safety assessments of nuclear power plants for the hypothetical large civil aircraft crash should be made mandatory, studies on large aircraft-nuclear power plant impact analyses and assessments are actively in progress. The large civil aircraft are being operated with a large amount of fuel and the fuel can be assumed to contribute to the impact loads at the impact. The fuel, i.e., the internal liquid can be considered as added masses classically in the evaluation of the impact load. According to the recent experimental research, it has been shown that the impact load of high speed impacting body with internal liquid is much higher than that of the mass-equivalent impacting body. In this study, the impact loads according to the existence of the internal liquid are computed by numerical methods and the safety assessment of nuclear power plant crashed by large civil aircraft are performed as an application

  9. Preliminary Data on the Safety of Phytoene- and Phytofluene-Rich Products for Human Use including Topical Application

    Directory of Open Access Journals (Sweden)

    Fabien Havas

    2018-01-01

    Full Text Available The colorless carotenoids phytoene and phytofluene are comparatively understudied compounds found in common foods (e.g., tomatoes and in human plasma, internal tissues, and skin. Being naturally present in common foods, their intake at dietary levels is not expected to present a safety concern. However, since the interest in these compounds in the context of many applications is expanding, it is important to conduct studies aimed at assessing their safety. We present here results of in vitro cytotoxicity and genotoxicity studies, revealing no significant cytotoxic or genotoxic potential and of short- and long-term human in vivo skin compatibility studies with phytoene- and phytofluene-rich tomato and Dunaliella salina alga extracts, showing a lack of irritancy or sensitization reactions. These results support the safe use of phytoene- and phytofluene-rich products in human topical applications.

  10. Safety evaluation report related to the preliminary design of the Standard Nuclear Steam Supply Reference System, RESAR SP/90 (Docket No. STN 50-601)

    International Nuclear Information System (INIS)

    1991-04-01

    On October 24, 1983, the Westinghouse Electric Corporation tendered its application for a preliminary design approval of the advanced pressurized-water reactor design for the SP/90 reactor. The Westinghouse Reference Safety Analysis Report (RESAR SP/90, Docket No. STN 50-601), describing the design of the facility, was submitted from October 24, 1983 through March 9, 1987. Staff of the US Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, has prepared this safety evaluation report of the RESAR SP/90 on the basis of its review. Because of the stage of the design, there are open issues that have not been resolved. These issues are discussed in detail throughout this report, and a summary is provided in Section 1.6 of this report. The applicant will be required to address these and any additional such concerns that may be raised during the course of the staff's review of advanced light-water reactors in support of a final design approval application. This report shall not constitute a commitment to issue a permit or license or in any way affect the authority of the Commission, its adjudicatory boards, and other presiding officers in any proceeding under Subpart G of Title 10 of the Code of Federal Regulations, Part 2

  11. [Preliminary results of an open-label observational study evaluating the efficacy and safety of Prolia used in women with postmenopausal osteoporosis].

    Science.gov (United States)

    Ershova, O B; Lesniak, O M; Belova, K Iu; Nazarova, A V; Manovitskaia, A V; Musaeva, T M; Musraev, R M; Nurlygaianov, R Z; Rozhinskaia, L Ia; Skripnikova, I A; Toroptsova, N V

    2014-01-01

    To evaluate the efficacy and safety of Denosumab (Prolia), a first-line osteoporosis (OP) medication that is a fully human monoclonal antibody to the receptor activator of nuclear factor xB ligand (RANKL), within an open-label observational study. Patients aged 50 years or older with postmenopausal OP, who were treated with Prolia in clinical practice, were examined. The concentrations of the bone resorption (BR) marker of C-terminal telopeptide and other laboratory indicators (total serum calcium, total alkaline phosphatase, and creatinine) were measured following 3 months. Adverse drug reactions were recorded. Three months after initiation of the investigation, there was a significant decrease in the BR marker C-terminal telopeptide (by 89%; p<0.0001). There were rare adverse reactions: hypocalcemia in 3 (5.9%) patients, arthralgias in 2 (3.9%), and eczema in 1 (1.9%). There were neither serious adverse events nor study withdrawal cases. The preliminary results of the open-label study of Prolia in postmenopausal OP suggest that the significantly lower BR activity determines the efficacy of this drug and its high safety.

  12. [European Union Network for Patient Safety and Quality of Care (PASQ). Development and preliminary results in Europe and in the Spanish National Health System].

    Science.gov (United States)

    Agra-Varela, Y; Fernández-Maíllo, M; Rivera-Ariza, S; Sáiz-Martínez-Acitorez, I; Casal-Gómez, J; Palanca-Sánchez, I; Bacou, J

    2015-01-01

    The joint action, European Union Network for Patient Safety and Quality of Care: PaSQ, aims to promote patient safety (PS) in the European Union (EU) and to facilitate the exchange of experiences among Member States (MS) and stakeholders on issues related to quality of care, PS, and patient involvement. The development and preliminary results are presented here, especially as regards the Spanish National Health System (SNHS). PaSQ is developed through 7 work packages, primarily aimed at sharing good practices (GP), which were identified using specific questionnaires and selected by means of explicit criteria, as well as to implement safe clinical practices (SCP) of proven effectiveness and agreed among MS. A total of 482 GP (39% provided by Spanish professionals) were identified. The 34 events organised in the EU, 11 including Spanish participation, facilitate sharing these practices. A total of 194 Health Care centres (49% in Spain) are implementing SCP (hand hygiene, safe surgery, medication reconciliation, and paediatric early warning scores) ACHIEVEMENTS AND FUTURE PERSPECTIVES: PaSQ is making it possible to strengthen collaboration between organizations and professionals at EU and SNHS level regarding PS and quality of care. Copyright © 2015 SECA. Published by Elsevier Espana. All rights reserved.

  13. Pharmacokinetics and Preliminary Safety of Pod-Intravaginal Rings Delivering the Monoclonal Antibody VRC01-N for HIV Prophylaxis in a Macaque Model.

    Science.gov (United States)

    Zhao, Chunxia; Gunawardana, Manjula; Villinger, Francois; Baum, Marc M; Remedios-Chan, Mariana; Moench, Thomas R; Zeitlin, Larry; Whaley, Kevin J; Bohorov, Ognian; Smith, Thomas J; Anderson, Deborah J; Moss, John A

    2017-07-01

    The broadly neutralizing antibody (bNAb) VRC01, capable of neutralizing 91% of known human immunodeficiency virus type 1 (HIV-1) isolates in vitro , is a promising candidate microbicide for preventing sexual HIV infection when administered topically to the vagina; however, accessibility to antibody-based prophylactic treatment by target populations in sub-Saharan Africa and other underdeveloped regions may be limited by the high cost of conventionally produced antibodies and the limited capacity to manufacture such antibodies. Intravaginal rings of the pod design (pod-IVRs) delivering Nicotiana -manufactured VRC01 (VRC01-N) over a range of release rates have been developed. The pharmacokinetics and preliminary safety of VRC01-N pod-IVRs were evaluated in a rhesus macaque model. The devices sustained VRC01-N release for up to 21 days at controlled rates, with mean steady-state VRC01-N levels in vaginal fluids in the range of 10 2 to 10 3 μg g -1 being correlated with in vitro release rates. No adverse safety indications were observed. These findings indicate that pod-IVRs are promising devices for the delivery of the candidate topical microbicide VRC01-N against HIV-1 infection and merit further preclinical evaluation. Copyright © 2017 American Society for Microbiology.

  14. Safety and preliminary evidence of biologic efficacy of a mammaglobin-a DNA vaccine in patients with stable metastatic breast cancer.

    Science.gov (United States)

    Tiriveedhi, Venkataswarup; Tucker, Natalia; Herndon, John; Li, Lijin; Sturmoski, Mark; Ellis, Matthew; Ma, Cynthia; Naughton, Michael; Lockhart, A Craig; Gao, Feng; Fleming, Timothy; Goedegebuure, Peter; Mohanakumar, Thalachallour; Gillanders, William E

    2014-12-01

    Mammaglobin-A (MAM-A) is overexpressed in 40% to 80% of primary breast cancers. We initiated a phase I clinical trial of a MAM-A DNA vaccine to evaluate its safety and biologic efficacy. Patients with breast cancer with stable metastatic disease were eligible for enrollment. Safety was monitored with clinical and laboratory assessments. The CD8 T-cell response was measured by ELISPOT, flow cytometry, and cytotoxicity assays. Progression-free survival (PFS) was described using the Kaplan-Meier product limit estimator. Fourteen subjects have been treated with the MAM-A DNA vaccine and no significant adverse events have been observed. Eight of 14 subjects were HLA-A2(+), and the CD8 T-cell response to vaccination was studied in detail. Flow cytometry demonstrated a significant increase in the frequency of MAM-A-specific CD8 T cells after vaccination (0.9% ± 0.5% vs. 3.8% ± 1.2%; P cells (41 ± 32 vs. 215 ± 67 spm; P cell responses, and preliminary evidence suggests improved PFS. Additional studies are required to define the potential of the MAM-A DNA vaccine for breast cancer prevention and/or therapy. ©2014 American Association for Cancer Research.

  15. Creation and preliminary validation of the screening for self-medication safety post-stroke scale (S-5).

    Science.gov (United States)

    Kaizer, Franceen; Kim, Angela; Van, My Tram; Korner-Bitensky, Nicol

    2010-03-01

    Patients with stroke should be screened for safety prior to starting a self-medication regime. An extensive literature review revealed no standardized self-medication tool tailored to the multi-faceted needs of the stroke population. The aim of this study was to create and validate a condition-specific tool to be used in screening for self-medication safety in individuals with stroke. Items were generated using expert consultation and review of the existing tools. The draft tool was pilot-tested on expert stroke clinicians to receive feedback on content, clarity, optimal cueing and domain omissions. The final version was piloted on patients with stroke using a structured interviewer-administered interview. The tool was progressively refined and validated according to feedback from the 11 expert reviewers. The subsequent version was piloted on patients with stroke. The final version includes 16 questions designed to elicit information on 5 domains: cognition, communication, motor, visual-perception and, judgement/executive function/self-efficacy. The Screening for Safe Self-medication post-Stroke Scale (S-5) has been created and validated for use by health professionals to screen self-medication safety readiness of patients after stroke. Its use should also help to guide clinicians' recommendations and interventions aimed at enhancing self-medication post-stroke.

  16. MIDAS documentation

    International Nuclear Information System (INIS)

    Seeman, S.E.

    1981-09-01

    MIDAS, the Master Information and Data Acquisition System, is a computerized system being implemented on FFTF to control the work released to the plant. The purpose of this system is to demonstrate the safety enhancement provided for LMFBRs when the operator has instant recall to the status of all work released to the plant, the interrelationships between functional equipment groups in the plant, and the relationships of equipment to safety functions

  17. Documentation of pediatric drug safety in manufacturers' product monographs: a cross-sectional evaluation of the canadian compendium of pharmaceuticals and specialities.

    Science.gov (United States)

    Uppal, Navjeet K; Dupuis, Lee L; Parshuram, Christopher S

    2008-01-01

    To describe the provision of pediatric drug safety information in a national formulary of manufacturers' drug product monographs. We performed a cross-sectional evaluation of comprehensive product monographs contained in the 2005 Canadian Compendium of Pharmaceuticals and Specialities (CPS). We abstracted data describing indications for prescription, statements about pediatric safety, available preparations, and provision of dosing guidelines. For each monograph we classified pediatric safety data as either present, present but limited or absent. We then described the pediatric safety data in CPS monographs for drugs listed in the published formulary of the Hospital for Sick Children, Toronto, Ontario, Canada. A total of 2232 product monographs were screened; 684 were excluded and 1548 (66%) were further analyzed. 1462 (94%) had indications that did not exclude children. Pediatric safety information was present in 592 (38%), present but limited in 148 (10%), and absent in 808 (52%) drug monographs. Safety statements were absent in 224 (14%) drug monographs that provided both dosing guidelines and formulations suitable for administration to children, and in 214 (52%) of 411 drugs in the pediatric hospital formulary. We evaluated a widely available national source of pediatric prescribing information. Safety data for children was not mentioned in more than half of the product monographs. Moreover, the provision of safety data was discordant with indications for prescription, the availability of pediatric formulations, and dosing guidelines within the monographs, and with inclusion in a pediatric hospital formulary. Our study suggests that the presentation of pediatric safety data in drug product monographs can be improved to better inform prescribing and to optimize pharmacotherapy in children.

  18. Safety and Effectiveness of Highly Active Antiretroviral Therapy in Treatment-Naïve HIV Patients: Preliminary Findings of a Cohort Event Monitoring Study in Belarus.

    Science.gov (United States)

    Setkina, Svetlana; Dotsenko, Marina; Bondar, Sviatlana; Charnysh, Iryna; Kuchko, Alla; Kaznacheeva, Alena; Kozorez, Elena; Dodaleva, Alena; Rossa, Natalia

    2015-04-01

    Antiretroviral drugs have well-documented evidence-based favorable benefit-risk ratios. Although various studies have investigated and characterized the safety profile of antiretroviral medicines, there are a limited number of studies evaluating the safety of first-line antiretroviral therapy (ART) in patients with a specific co-morbidity. A cohort event monitoring (CEM) study of the safety and effectiveness of antiretroviral medicines in a target population that has a significant level of co-morbidities (chronic infectious diseases, peripheral blood cytopenias) was implemented. The aim was to evaluate the safety profile of the highly active ART (HAART) in the target population and subpopulations with risk factors, to optimize the monitoring and decision-making procedure for subgroups of patients with specific types of co-morbidity, and to implement a more vigilant approach to therapy management in risk groups of patients. Prospective observational CEM was implemented among HAART-naïve HIV-positive patients at four clinical sites from December 2012. Eligible patients were those starting first-line HAART. Close medical supervision of all enrolled patients, with regular clinical and laboratory monitoring, was provided by healthcare professionals within 1 year after commencement of therapy. Standardized forms were used for data collection on initial and subsequent visits. All objective or subjective deviations in condition (events) were assessed for a causal relationship with ART, and for severity, seriousness, reversibility, preventability, and pre-existing risk factors in the case of adverse drug reactions (ADRs). A total of 518 HAART-naïve HIV-positive patients were enrolled in the CEM study. Of these patients, 65% (337) experienced one or several ADRs related to one or more components of HAART. Most of the ADRs reported were non-serious, expected, common (very common), transient (correctable), or reversible. The most common were hematotoxic, hepatotoxic, and

  19. Feasibility and Safety of a Powered Exoskeleton for Assisted Walking for Persons With Multiple Sclerosis: A Single-Group Preliminary Study.

    Science.gov (United States)

    Kozlowski, Allan J; Fabian, Michelle; Lad, Dipan; Delgado, Andrew D

    2017-07-01

    To examine the feasibility, safety, and secondary benefit potential of exoskeleton-assisted walking with one device for persons with multiple sclerosis (MS). Single-group longitudinal preliminary study with 8-week baseline, 8-week intervention, and 4-week follow-up. Outpatient MS clinic, tertiary care hospital. Participants (N=13; age range, 38-62y) were mostly women with Expanded Disability Status Scale scores ranging from 5.5 to 7.0. Exoskeleton-assisted walk training. Primary outcomes were accessibility (enrollment/screen pass), tolerability (completion/dropout), learnability (time to event for standing, walking, and sitting with little or no assistance), acceptability (satisfaction on the device subscale of the Quebec User Evaluation of Satisfaction with Assistive Technology version 2), and safety (event rates standardized to person-time exposure in the powered exoskeleton). Secondary outcomes were walking without the device (timed 25-foot walk test and 6-minute walk test distance), spasticity (Modified Ashworth Scale), and health-related quality of life (Patient-Reported Outcomes Measurement and Information System pain interference and Quality of Life in Neurological Conditions fatigue, sleep disturbance, depression, and positive affect and well-being). The device was accessible to 11 and tolerated by 5 participants. Learnability was moderate, with 5 to 15 sessions required to walk with minimal assistance. Safety was good; the highest adverse event rate was for skin issues at 151 per 1000 hours' exposure. Acceptability ranged from not very satisfied to very satisfied. Participants who walked routinely improved qualitatively on sitting, standing, or walking posture. Two participants improved and 2 worsened on ≥1 quality of life domain. The pattern of spasticity scores may indicate potential benefit. The device appeared feasible and safe for about a third of our sample, for whom routine exoskeleton-assisted walking may offer secondary benefits. Copyright

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Preliminary safety and efficacy results with robotic high-intensity focused ultrasound : A single center Indian experience

    Directory of Open Access Journals (Sweden)

    Shashikant Mishra

    2011-01-01

    Full Text Available Background : There are no Indian data of high-intensity focused ultrasound (HIFU. Being an alternative, still experimental modality, reporting short-term safety outcome is paramount. Aims : This study was aimed at to assess the safety and short-term outcome in patients with prostate cancer treated by HIFU. Settings and Design : A retrospective study of case records of 30 patients undergoing HIFU between January 2008 to September 2010 was designed and conducted. Materials and Methods : The procedural safety was analyzed at 3 months. Follow-up consisted of 3 monthly prostate-specific antigen (PSA levels and transrectal biopsy if indicated. All the patients had a minimum follow-up of 6 months. Results : A mean prostate volume of 26.9 ± 8.5 cm 3 was treated in a mean time of 115 ± 37.4 min. There was no intraoperative complication. The postoperative pain visual analogue score at day 0 was 2.1 ± 1.9 and at day 1 was 0.4 ± 0.8 on a scale of 1-10. Mean duration of perurethral catheter removal was 3.9 days. The complications after treatment were: LUTS in seven patients, stress incontinence in two, stricture in two, and symptomatic urinary tract infection in five. Average follow-up duration was 10.4 months (range, 6-20 months. Mean time to obtain PSA nadir was 6 ± 3 months with a median PSA nadir value of 0.3 ng/ml. Two patients had positive prostatic biopsy in the localized (high risk group. Conclusions : HIFU was safe in carcinoma prostate patients. The short-term results were efficacious in localized disease. The low complication rates and favorable functional outcome support the planning of further larger studies.

  2. Preliminary safety analysis of criticality for dual-purpose metal cask under dry storage conditions in South Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taeman, E-mail: tmkim@korad.or.kr [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Dho, Hoseog; Baeg, Chang-Yeal [Korea Radioactive Waste Agency (KORAD), 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Lee, Gang-uk [Korea Nuclear Engineering and Service Co. (KONES), Hyundai Plaza, 341-4 Jangdae-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2014-10-15

    Highlights: • DPC is under development led by Korea Radioactive Waste Agency in South Korea. • The results of criticality analysis with respect to design requirements. • The k{sub eff} under normal and off-normal conditions were 0.36 and 0.46, respectively. • In addition, the k{sub eff} under a postulated accident condition was evaluated to be 0.94. - Abstract: A dual-purpose metal cask is under development led by Korea Radioactive Waste Agency (KORAD) in Korea, for the dry interim storage and long-distance transportation. This cask comprises a main body made of carbon steel and a stainless steel Dry Shielded Canister (DSC), with stainless steel baskets inside to contain spent fuel assemblies. In this study, nuclear criticality safety analysis was conducted as a part of safety assessment of the metal cask. Analysis to show criticality safety in accordance with regulatory requirements of PWR spent fuel storage was carried out. 10CFR72.124 “Criteria for nuclear criticality safety” and the Regulatory Guide of the American Nuclear Society, ANSI/ANS-57.9 “Design Criteria for an Independent Spent Fuel” and US NRC's “Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility” were employed as regulatory standard and criteria. This paper shows results of criticality analysis with respect to each designated criterion with modeling of a virtual nuclear fuel assembly and a cask body that induces the maximum reactivity among various design basis fuels of the metal cask. In addition, the sensitivity analysis of nuclear criticality taking into account the various modeling deviation such as manufacturing tolerance and modeling assumptions of conventional models was carried out to ensure the reliability of the analysis result. The criticality evaluation result of the metal cask and the maximum k{sub eff} under normal and off-normal conditions were 0.36884 and 0.46255, respectively. The maximum k{sub eff} under a postulated

  3. Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant; Resultados preliminares mas significativos del analysis probabilista de seguridad de la Central Nuclear de Juragua

    Energy Technology Data Exchange (ETDEWEB)

    Perdomo, Manuel [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

    1995-12-31

    Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential `weak points` at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs.

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

    International Nuclear Information System (INIS)

    Scarlat, Raluca O.; Cisneros, Anselmo T.; Koutchesfahani, Tawni; Hong, Rada; Peterson, Per F.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  6. CNEA's quality system documentation

    International Nuclear Information System (INIS)

    Mazzini, M.M.; Garonis, O.H.

    1998-01-01

    Full text: To obtain an effective and coherent documentation system suitable for CNEA's Quality Management Program, we decided to organize the CNEA's quality documentation with : a- Level 1. Quality manual. b- Level 2. Procedures. c-Level 3. Qualities plans. d- Level 4: Instructions. e- Level 5. Records and other documents. The objective of this work is to present a standardization of the documentation of the CNEA's quality system of facilities, laboratories, services, and R and D activities. Considering the diversity of criteria and formats for elaboration the documentation by different departments, and since ultimately each of them generally includes the same quality management policy, we proposed the elaboration of a system in order to improve the documentation, avoiding unnecessary time wasting and costs. This will aloud each sector to focus on their specific documentation. The quality manuals of the atomic centers fulfill the rule 3.6.1 of the Nuclear Regulatory Authority, and the Safety Series 50-C/SG-Q of the International Atomic Energy Agency. They are designed by groups of competent and highly trained people of different departments. The normative procedures are elaborated with the same methodology as the quality manuals. The quality plans which describe the organizational structure of working group and the appropriate documentation, will asses the quality manuals of facilities, laboratories, services, and research and development activities of atomic centers. The responsibilities for approval of the normative documentation are assigned to the management in charge of the administration of economic and human resources in order to fulfill the institutional objectives. Another improvement aimed to eliminate unnecessary invaluable processes is the inclusion of all quality system's normative documentation in the CNEA intranet. (author) [es

  7. Use of site specific data from Aespoe - preliminary results from the on-going safety analysis SR 97

    International Nuclear Information System (INIS)

    Stroem, A.; Selroos, J.O.; Andersson, Johan

    1998-01-01

    This paper discusses an on-going safety assessment study of SKB as well as the use of field data from Aespoe for obtaining input parameters for flow and radionuclide transport modelling in the geosphere. In the on-going Safety Assessment study SR 97, three individual sites in Sweden are used for exemplifying site specific conditions on overall repository performance. Thus, models capable of reproducing site specific characteristics are utilised. This is primarily obtained by implementing the geologic structural models in suitable conceptual models for groundwater flow on both regional and local scales. The models for flow incorporate observed and/or inferred water conducting features as well as other site-specific characteristics necessary for realistic descriptions of flow at the sites. The flow modelling thus aims at realism; the results obtained for present day conditions should not in any serious aspect conflict with observations at the site. Agreement between observed and modelled entities provides confidence in that a sound understanding of the site is obtained. Aespoe is one of the three sites providing site-specific conditions in SR 97. Transport is subsequently modelled using a stream tube approach where the 'travel times', for non-sorbing species, and discharge locations of a set of one-dimensional stream tubes are obtained from particle tracking in the flow model. The resulting distribution of 'travel times' in a single model realisation reflects the spatial variability and spatial extent of the repository, whereas the ensemble travel time distribution (over several realisations) for a given canister location reflects the uncertainty in travel time. The actual transport paths used in the transport modelling are thus dependent on site specific information such as e.g. existence of water conductive features. Other input parameters to the transport model are based on more generic and/or conservative arguments. However, the goal in a safety assessment

  8. FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    KRIPPS, L.J.

    2005-03-03

    This document describes the qualitative evaluation of frequency and consequences for DST and SST representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant structures, systems and components (SSCs) and/or technical safety requirements (TSRs) were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support WP-13033, Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

  9. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    International Nuclear Information System (INIS)

    Heard, F.J.

    1997-01-01

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

  10. Safety assessment of radioactive waste disposal into geological formations; a preliminary application of fault tree analysis to salt deposits

    International Nuclear Information System (INIS)

    Bertozzi, B.; D'Alessandro, M.; Girardi, F.; Vanossi, M.

    1978-01-01

    The methodology of the fault tree analysis (FTA) has been widely used at the Joint Research Centre of Ispra in nuclear reactor safety studies. The aim of the present work consisted in studying the applicability of this methodology to geological repositories of radioactive wastes, including criteria and approaches for the quantification of probalities of primary events. The present work has just an illustrative purpose. Two ideal cases of saline formations, I.E. a bedded salt and a diapir were chosen as potential disposal sites for radioactive waste. On the basis of arbitrarily assumed hydrogeological features of the salt formations and their surrounding environment, possible phenomena capable of causing the waste to be released from each formation have been discussed and gathered following the logical schemes of the FTA. The assessment of probability values for release events due to natural causes as well as to human actions, over different time periods, up to one million years, has been discussed

  11. A Disposable Tear Glucose Biosensor-Part 4: Preliminary Animal Model Study Assessing Efficacy, Safety, and Feasibility.

    Science.gov (United States)

    La Belle, Jeffrey T; Engelschall, Erica; Lan, Kenneth; Shah, Pankti; Saez, Neil; Maxwell, Stephanie; Adamson, Teagan; Abou-Eid, Michelle; McAferty, Kenyon; Patel, Dharmendra R; Cook, Curtiss B

    2014-01-01

    A prototype tear glucose (TG) sensor was tested in New Zealand white rabbits to assess eye irritation, blood glucose (BG) and TG lag time, and correlation with BG. A total of 4 animals were used. Eye irritation was monitored by Lissamine green dye and analyzed using image analysis software. Lag time was correlated with an oral glucose load while recording TG and BG readings. Correlation between TG and BG were plotted against one another to form a correlation diagram, using a Yellow Springs Instrument (YSI) and self-monitoring of blood glucose as the reference measurements. Finally, TG levels were calculated using analytically derived expressions. From repeated testing carried over the course of 12 months, little to no eye irritation was detected. TG fluctuations over time visually appeared to trace the same pattern as BG with an average lag times of 13 minutes. TG levels calculated from the device current measurements ranged from 4 to 20 mg/dL and correlated linearly with BG levels of 75-160 mg/dL (TG = 0.1723 BG = 7.9448 mg/dL; R 2 = .7544). The first steps were taken toward preliminary development of a sensor for self-monitoring of tear glucose (SMTG). No conjunctival irritation in any of the animals was noted. Lag time between TG and BG was found to be noticeable, but a quantitative modeling to correlate lag time in this study is unnecessary. Measured currents from the sensors and the calculated TG showed promising correlation to BG levels. Previous analytical bench marking showed BG and TG levels consistent with other literature. © 2014 Diabetes Technology Society.

  12. Safety of cryopreservation straws for human gametes or embryos: a preliminary study with human immunodeficiency virus-1.

    Science.gov (United States)

    Benifla, J L; Letur-Konïrsch, H; Collin, G; Devaux, A; Kuttenn, F; Madelenat, P; Brun-Vezinet, F; Feldmann, G

    2000-10-01

    The aim of this preliminary experimental study was to test the stability of cryopreservation straws to human immunodeficiency virus-1 (HIV-1). Three kinds of straws were tested: four polyvinyl chloride (PVC), four polyethylene terephthalate glycol (PETG) and 20 high-security ionomeric resin (IR). The PVC and PETG straws were sealed ultrasonically, and the IR straw by thermosoldering. Each sealed straw was cut in half to produce two demi-straws and then filled with 100 microl of HIV-1-containing supernatant (reverse transcriptase activity: 15 000 c.p.m./50 microl). The unsealed cotton end of PVC and PETG straws and the two halves of the IR straws (cotton and plastic plug ends) were tested. Each demi-straw was two- thirds submerged in RPMI medium at 37 degrees C, and RPMI samples were withdrawn on days 3, 7 and 11. Viral RNA was extracted from the medium and then amplified by reverse transcriptase-polymerase chain reaction (RT-PCR) followed by nested PCR using primers specific to HIV-1 protease. On day 7, no HIV-1 RNA was detected in any of the different samples of medium that had surrounded the unsealed PVC and PETG straws with cotton ends, but three IR specimens were positive. On day 11, PVC and PETG remained negative but HIV-1 RNA was detected in RPMI samples for two more IR demi-straws (n = 5). In conclusion, under these experimental conditions (at 37 degrees C), the unsealed cotton end PVC, PETG and thermosoldered cotton end IR demi-straws appeared to be safe for HIV-1, while IR straws, sealed or unsealed with a plastic plug and with unsealed cotton ends, leaked.

  13. On the Issue of Regional Policy in the Sphere of Ecological Safety (Based on the Documents Concerning the Republic of Kalmykia

    Directory of Open Access Journals (Sweden)

    Elza I. Mantaeva

    2018-03-01

    Full Text Available The regional government should provide ecological safety in order to ensure stable social and economic development of the country. The article outlines theoretical approach and recommendations on practice concerning ecological issues, investigates state policy aiming to preserve nature and analyses sources of threat to ecological safety in the region. The modern paradigm of the country’s and region’s development is characterized by the combination of ecological and economic policy. The ecological policy providing measures for efficient use of natural resources, their protection and restoration is implemented in the economic strategy of highly developed countries and encourages stable social and economic development of the country and the region. Some researchers assert that Russia’s most regions do not possess sufficient resources necessary to implement structural shifts which can eliminate the main threats to ecological safety. The Republic of Kalmykia can be included into this list. It is rich in natural resources but they are not restored due to the lack of regulation for efficient consumption. The main concerns of the republic are water supply, desertification, and pollution of atmosphere, waste disposal, and protection of wildlife and environment, preservation of hunting ground, forest management. The republic generates no electricity so it is delivered by neighboring regions. The republic should use the renewable ways of producing energy which can contribute to the energy and ecological safety and can attract new investment funds. There is an urgent need for stricter state regulation to ensure socioeconomic development and ecological safety in Kalmykia.

  14. Preliminary assessment of cardiac short term safety and efficacy of manganese chloride for cardiovascular magnetic resonance in humans

    Directory of Open Access Journals (Sweden)

    Kalaf Jose M

    2011-01-01

    Full Text Available Abstract Background Manganese based agents are intracellular and accumulate inside myocytes allowing for different imaging strategies compared to gadolinium contrasts. While previous agents release manganese very slowly in the circulation, MnCl2 allows for rapid Mn2+ uptake in myocytes, creating a memory effect that can be potentially explored. Data on animal models are very encouraging but the safety and efficacy of this approach in humans has not yet been investigated. Therefore, our objectives were to study the safety and efficacy of a rapid infusion of manganese chloride (MnCl2 for cardiovascular magnetic resonance (CMR in humans. Methods Fifteen healthy volunteers underwent a CMR scan on a 1.5 T scanner. Before the infusion, cardiac function was calculated and images of a short axis mid-ventricular slice were obtained using a 2D and 3D gradient-echo inversion recovery (GRE-IR sequence, a phase-sensitive IR sequence and a single breath-hold segmented IR prepared steady-state precession acquisition for T1 calculations. MnCl2 was infused over three minutes at a total dose of 5 μMol/kg. Immediately after the infusion, and at 15 and 30 minutes later, new images were obtained and cardiac function re-evaluated. Results There was a significant decrease in T1 values compared to baseline, sustained up to 30 minutes after the MnCl2 infusion (pre,839 ± 281 ms; 0 min, 684 ± 99; 15 min, 714 ± 168; 30 min, 706 ± 172, P = 0.003. The 2D and 3D GRE-IR sequence showed the greatest increase in signal-to-noise ratio compared to the other sequences (baseline 6.6 ± 4.2 and 9.7 ± 5.3; 0 min, 11.3 ± 4.1 and 15.0 ± 8.7; 15 min, 10.8 ± 4.0 and 16.9 ± 10.2; 30 min, 10.6 ± 5.2 and 16.5 ± 8.3, P 2 with no major adverse events, despite all reporting transient facial flush. Conclusions In the short term, MnCl2 appears safe for human use. It effectively decreases myocardium T1, maintaining this effect for a relatively long period of time and allowing for the

  15. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the iCMS Web site. The following items can be found on: http://cms.cern.ch/iCMS/ General - CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. LHC Symposiums Management - CB - MB - FB - FMC Agendas and minutes are accessible to CMS members through their AFS account (ZH). However some linked documents are restricted to the Board Members. FB documents are only accessible to FB members. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2006 Annual reviews are posted.   CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral students upon completion of their theses. Therefore it is requested that Ph.D students inform the CMS Secretariat a...

  16. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the iCMS Web site. The following items can be found on: http://cms.cern.ch/iCMS/ General - CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. LHC Symposiums Management - CB - MB - FB - FMC Agendas and minutes are accessible to CMS members through their AFS account (ZH). However some linked documents are restricted to the Board Members. FB documents are only accessible to FB members. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2006 Annual reviews are posted. CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral students upon completion of their theses. Therefore it is requested that Ph.D students inform the CMS Secretariat about the natu...

  17. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the iCMS Web site. The following items can be found on: http://cms.cern.ch/iCMS/ General - CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. LHC Symposiums Management - CB - MB - FB - FMC Agendas and minutes are accessible to CMS members through their AFS account (ZH). However some linked documents are restricted to the Board Members. FB documents are only accessible to FB members. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2006 Annual reviews are posted. CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral students upon completion of their theses. Therefore it is requested that Ph.D students inform the CMS Secretariat about the natur...

  18. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the iCMS Web site. The following items can be found on: http://cms.cern.ch/iCMS/ Management- CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. Management - CB - MB - FB Agendas and minutes are accessible to CMS members through their AFS account (ZH). However some linked documents are restricted to the Board Members. FB documents are only accessible to FB members. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2007 Annual reviews are posted. CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral students upon completion of their theses. Therefore it is requested that Ph.D students inform the CMS Secretariat about the nature of employment and ...

  19. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the iCMS Web site. The following items can be found on: http://cms.cern.ch/iCMS/ Management- CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. Management - CB - MB - FB Agendas and minutes are accessible to CMS members through their AFS account (ZH). However some linked documents are restricted to the Board Members. FB documents are only accessible to FB members. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2007 Annual reviews are posted. CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral students upon completion of their theses. Therefore it is requested that Ph.D students inform the CMS Secretariat about the nature of em¬pl...

  20. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the iCMS Web site. The following items can be found on: http://cms.cern.ch/iCMS/ General - CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. LHC Symposiums Management - CB - MB - FB - FMC Agendas and minutes are accessible to CMS members through their AFS account (ZH). However some linked documents are restricted to the Board Members. FB documents are only accessible to FB members. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2006 Annual reviews are posted. CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral students upon completion of their theses. Therefore it is requested that Ph.D students inform the CMS Secretariat about the na...

  1. A preliminary assessment of two hospital food service systems using parameters of food safety and consumer opinion.

    Science.gov (United States)

    Hartwell, H; Edwards, J S

    2001-12-01

    The goal of any hospital caterer should be to provide food that meets nutritional requirements, satisfies the patient, improves morale and is microbiologically safe. Food distribution to hospital wards plays a critical role. The aim of this study was to compare two hospital food service systems using parameters of food safety and consumer opinion. An NHS hospital was selected where food delivery was due to change from a plated system to a cafeteria trolley system. Samples (50 g) of dishes (n = 27) considered to be high-risk were collected for three consecutive days from breakfast, lunch and supper meals. The samples were taken from a pre-ordered tray (similar to that of a patient) in the plated system and from the trolley on the ward in the cafeteria system of meal delivery (approximately six months after its introduction). Consumer opinions cards (n = 180) were distributed and interviews also conducted. Microbiologically, the quality of food items delivered by both systems was satisfactory. However, concern was raised with the plated system, not for hot foods cooling down but for chilled foods warming up and being sustained in ambient conditions. Overall consumer satisfaction and experience was enhanced with the trolley system. Food was hotter and generally perceived to be of a better quality. Satisfaction with cold desserts was not dependent on the delivery system.

  2. Safety and preliminary immunogenicity of Cuban pneumococcal conjugate vaccine candidate in healthy children: a randomized phase I clinical trial.

    Science.gov (United States)

    Dotres, Carlos P; Puga, Rinaldo; Ricardo, Yariset; Broño, Carmen R; Paredes, Beatriz; Echemendía, Vladimir; Rosell, Sandra; González, Nadezhda; García-Rivera, Dagmar; Valdés, Yury; Goldblatt, David; Vérez-Bencomo, Vicente

    2014-09-15

    A new heptavalent conjugate vaccine (PCV7-TT) is under development in Cuba. PCV7-TT contains 2 μg of serotypes 1, 5, 14, 18C, 19F, 23F and 4 μg of 6B, each one conjugated to tetanus toxoid (TT). This vaccine was designed with the serotypes that cause most invasive pneumococcal diseases (IPD) worldwide. In the present study, we investigated the safety and explored the immunogenicity of PCV7-TT during a controlled, randomized and double blind clinical trial phase I in 4-5-year-old children. PCV7-TT was well tolerated and as safe as Synflorix used as control vaccine. Following a single-dose vaccination, all individual serotypes included in PCV7-TT induced statistically significant increase of IgG GMC and OPA GMT. These are the first clinical results of PCV7-TT in children and they pave the way toward next clinical trials in children and infants. This clinical trial was published in the Cuban Public Register of Clinical Trials with code RPCEC00000173. Copyright © 2014 Elsevier Ltd. All rights reserved.

  3. A preliminary case series evaluating the safety and immediate to short-term clinical benefits of joint mobilization in hemophilic arthritis of the lower limb.

    Science.gov (United States)

    Scaddan, Emma; Rowell, John; O'Leary, Shaun

    2017-09-01

    Arthritis resulting from recurrent intra-articular bleeding in individuals with hemophilia can be severely debilitating due to joint pain and stiffness with subsequent loss of mobility and function. Very limited studies have investigated the potential benefits of joint mobilization for this condition. This case series is a preliminary investigation of safety, as well as immediate and short-term clinical benefits, associated with gentle knee and ankle joint mobilization in people with hemophilic arthropathy. A single intervention of joint mobilization was applied to the affected knees and/or ankles of 16 individuals with severe or moderate hemophilia within a public hospital setting. Adverse events, as well as immediate (pain-free passive joint range, Timed Up and Go Test with maximum pain numerical rating scale) and short-term (Lower Extremity Functional Scale) effects of the intervention were evaluated with a repeated measures ANOVA. There were no adverse events. An immediate significant increase was observed in pain-free passive ankle joint range of motion ( p  < 0.05) following the joint mobilization intervention. The findings of this case series suggest that gentle joint mobilization techniques may be safely considered as part of a multimodal management approach for hemophilic arthropathy.

  4. Intraarterial reteplase and intravenous abciximab for treatment of acute ischemic stroke. A preliminary feasibility and safety study in a non-human primate model

    International Nuclear Information System (INIS)

    Qureshi, Adnan I.; Suri, M. Fareed K.; Ali, Zulfiqar; Ringer, Andrew J.; Boulos, Alan S.; Guterman, Lee R.; Hopkins, L. Nelson; Nakada, Marian T.; Alberico, Ronald A.; Martin, Lisa B.E.

    2005-01-01

    We performed a preliminary feasibility and safety study using intravenous (IV) administration of a platelet glycoprotein IIb/IIIa inhibitor (abciximab) in conjunction with intraarterial (IA) administration of a thrombolytic agent (reteplase) in a primate model of intracranial thrombosis. We introduced thrombus through superselective catheterization of the intracranial segment of the internal carotid artery in 16 primates. The animals were randomly assigned to receive IA reteplase and IV abciximab (n =4), IA reteplase and IV placebo (n =4), IA placebo and IV abciximab (n =4) or IA and IV placebo (n =4). Recanalization was assessed by serial angiography during the 6-h period after initiation of treatment. Postmortem magnetic resonance (MR) imaging was performed to determine the presence of cerebral infarction or intracranial hemorrhage. Partial or complete recanalization at 6 h after initiation of treatment (decrease of two or more points in pre-treatment angiographic occlusion grade) was observed in two animals treated with IA reteplase and IV abciximab, three animals treated with IA reteplase alone and one animal treated with IV abciximab alone. No improvement in perfusion was observed in animals that received IV and IA placebo. Cerebral infarction was demonstrated on postmortem MR imaging in three animals that received IA and IV placebo and in one animal each from the groups that received IA reteplase and IV abciximab or IV abciximab alone. One animal that received IV abciximab alone had a small intracerebral hemorrhage on MR imaging. (orig.)

  5. Nintendo Wii Fit as an adjunct to physiotherapy following lower limb fractures: preliminary feasibility, safety and sample size considerations.

    Science.gov (United States)

    McPhail, S M; O'Hara, M; Gane, E; Tonks, P; Bullock-Saxton, J; Kuys, S S

    2016-06-01

    The Nintendo Wii Fit integrates virtual gaming with body movement, and may be suitable as an adjunct to conventional physiotherapy following lower limb fractures. This study examined the feasibility and safety of using the Wii Fit as an adjunct to outpatient physiotherapy following lower limb fractures, and reports sample size considerations for an appropriately powered randomised trial. Ambulatory patients receiving physiotherapy following a lower limb fracture participated in this study (n=18). All participants received usual care (individual physiotherapy). The first nine participants also used the Wii Fit under the supervision of their treating clinician as an adjunct to usual care. Adverse events, fracture malunion or exacerbation of symptoms were recorded. Pain, balance and patient-reported function were assessed at baseline and discharge from physiotherapy. No adverse events were attributed to either the usual care physiotherapy or Wii Fit intervention for any patient. Overall, 15 (83%) participants completed both assessments and interventions as scheduled. For 80% power in a clinical trial, the number of complete datasets required in each group to detect a small, medium or large effect of the Wii Fit at a post-intervention assessment was calculated at 175, 63 and 25, respectively. The Nintendo Wii Fit was safe and feasible as an adjunct to ambulatory physiotherapy in this sample. When considering a likely small effect size and the 17% dropout rate observed in this study, 211 participants would be required in each clinical trial group. A larger effect size or multiple repeated measures design would require fewer participants. Copyright © 2015 Chartered Society of Physiotherapy. Published by Elsevier Ltd. All rights reserved.

  6. Clinical efficacy and safety evaluation of a novel fractional unipolar radiofrequency device on facial tightening: A preliminary report.

    Science.gov (United States)

    Suh, Dong Hye; Byun, Eun Jung; Lee, Sang Jun; Song, Kye Yong; Kim, Hei Sung

    2017-06-01

    Previous studies have shown that radiofrequency (RF) energy is safe and effective for improving skin laxity. Unlike monopolar and bipolar devices, little has been studied with the unipolar hand piece. We sought to evaluate the safety and efficacy of a novel fractional unipolar RF device on facial tightening. This was a retrospective, single-center study of 14 subjects with age-related facial laxity who underwent five sessions of fractional unipolar RF at an interval of 2 weeks, and then followed-up for 3 months. Standardized photos were taken at baseline and at 3-months follow-up, and were assessed by two independent dermatologists using a 4-point scale (0=no improvement, 1=mild improvement, 2=moderate improvement, 3=significant improvement). Punch biopsies (2 mm) were performed and a questionnaire was used to evaluate the patient's satisfaction and the incidence of adverse reactions. Fourteen subjects with mild to moderate age-related facial laxity were included in the study. The mean age of the subjects was 49.7 years (range 32-80). 35.7% of the subjects showed significant improvement, 50% moderate improvement, and 14.3% slight improvement of facial laxity in their follow-up photos. About 85.7% of the patients replied that they were either greatly satisfied or satisfied with the results at 3-months follow-up. Skin biopsies revealed an increase in collagen in the dermis. None of the subjects experienced any serious adverse events during or after the procedure. Our findings suggest that fractional Unipolar RF can be safely performed on the face and is effective in skin tightening. It has a great advantage over other forms of RF by being entirely painless. © 2017 Wiley Periodicals, Inc.

  7. Document Models

    Directory of Open Access Journals (Sweden)

    A.A. Malykh

    2017-08-01

    Full Text Available In this paper, the concept of locally simple models is considered. Locally simple models are arbitrarily complex models built from relatively simple components. A lot of practically important domains of discourse can be described as locally simple models, for example, business models of enterprises and companies. Up to now, research in human reasoning automation has been mainly concentrated around the most intellectually intensive activities, such as automated theorem proving. On the other hand, the retailer business model is formed from ”jobs”, and each ”job” can be modelled and automated more or less easily. At the same time, the whole retailer model as an integrated system is extremely complex. In this paper, we offer a variant of the mathematical definition of a locally simple model. This definition is intended for modelling a wide range of domains. Therefore, we also must take into account the perceptual and psychological issues. Logic is elitist, and if we want to attract to our models as many people as possible, we need to hide this elitism behind some metaphor, to which ’ordinary’ people are accustomed. As such a metaphor, we use the concept of a document, so our locally simple models are called document models. Document models are built in the paradigm of semantic programming. This allows us to achieve another important goal - to make the documentary models executable. Executable models are models that can act as practical information systems in the described domain of discourse. Thus, if our model is executable, then programming becomes redundant. The direct use of a model, instead of its programming coding, brings important advantages, for example, a drastic cost reduction for development and maintenance. Moreover, since the model is well and sound, and not dissolved within programming modules, we can directly apply AI tools, in particular, machine learning. This significantly expands the possibilities for automation and

  8. EVALUATION OF SAFETY AND EFFECTIVENESS OF HYLAN GF-20 IN PATIENTS WITH KNEE OSTEOARTHRITIS IN REAL LIFE PRACTICE: PRELIMINARY RESULTS

    Directory of Open Access Journals (Sweden)

    G. V. Kuropatkin

    2016-01-01

    Full Text Available Combined conservative treatment of osteoarthritis includes intra-articular injections of hyaluronic acid. The paper presents the interim results of a prospective observational multicenter non-comparative study conducted in accordance with routine clinical practice to assess the safety and effectiveness of Hylan GF-20 in patients with knee osteoarthritis. The primary objective of the study is to assess walking and rest pain severity by WOMAC VA3.1 scale after 26 weeks and 52 weeks compared to the baseline. To date, 42 patients completed the study (71.43% women, 28.57% men, patients mean age is 59.79 years. After intraarticular injection of Hylan GF-20 patients were examined after 3, 6 and 12 months. After 3 months a positive clinical response was observed: pain severity decreased by 51.31% (p<0,001 on WOMAC A, joint stiffness decreased by 51.02% (p<0,001 on WOMAC B, daily life difficulties decreased by 42.03% (p<0,001 on WOMAC C. The same tendency was observed in the following periods. By week 52 pain severity reduced by 53.25% (p<0,001 on WOMAC A, joint stiffness by 41.63% (p<0,001 on WOMAC B, daily life difficulties -by 47.55% (p<0.001 on WOMAC C. Level of clinical response didn't correlate with the osteoarthritis stage. Therapy resulted in improvement of life quality of patients with knee osteoarthritis by 36% (p<0,001 according to the questionnaire EQ-5D. The general patients status according to physicians improved by 33.96% (p<0,001 on VAS, and by 45.91% (p<0,001 according to patients response on VAS. By week 52 of follow-up 11,90% of patients demonstrated a decrease in required concomitant therapy of osteoarthritis. Given the chronic nature of the osteoarthritis it's the most important that the therapeutic effect of Hylan GF-20 maintains until week 52. The above results confirm the efficiency of intraar-ticular injection of hyaluronic acid in patients with knee osteoarthritis especially in early disease stage.

  9. Risk assessment. Basic documents for the preparation of the report 'Energy, Health, environment, and safety hazards' by the Swedish Energy Commission

    International Nuclear Information System (INIS)

    1978-01-01

    This report contains several essays on the rational ways of assessing the hazards, and choosing, the future energy production systems. Theories of decision analysis are touched upon by a number of authors and the fault tree statistical method is described. The Rasmussen report (WASH-1400) and a few related papers are discussed. As an example of the risks of advanced technological systems, the air traffic safety is accounted for. (L.E.)

  10. FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    KRIPPS, L.J.

    2005-02-18

    This document describes the qualitative evaluation of frequency and consequences for double shell tank (DST) and single shell tank (SST) representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant SSCs and/or TSRS were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support of the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

  11. ITER safety

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. Orbitmpi Documentation

    International Nuclear Information System (INIS)

    Lowe, Lisa L.

    2000-01-01

    Orbitmpi is a parallelized version of Roscoe White's Orbit code. The code has been parallelized using MPI, which makes it portable to many types of machines. The guidelines used for the parallelization were to increase code performance with minimal changes to the code's original structure. This document gives a general description of how the parallel sections of the code run. It discusses the changes made to the original code and comments on the general procedure for future additions to Orbitmpi, as well as describing the effects of a parallelized random number generator on the code's output. Finally, the scaling results from Hecate and from Puffin are presented. Hecate is a 64-processor Origin 2000 machine, with MIPS R12000 processors and 16GB of memory, and Puffin is a PC cluster with 9 dual-processor 450 MHz Pentium III (18 processors max.), with 100Mbits ethernet communication

  13. Nuclear power plants documentation system

    International Nuclear Information System (INIS)

    Schwartz, E.L.

    1991-01-01

    Since the amount of documents (type and quantity) necessary for the entire design of a NPP is very large, this implies that an overall and detailed identification, filling and retrieval system shall be implemented. This is even more applicable to the FINAL QUALITY DOCUMENTATION of the plant, as stipulated by IAEA Safety Codes and related guides. For such a purpose it was developed a DOCUMENTATION MANUAL, which describes in detail the before mentioned documentation system. Here we present the expected goals and results which we have to reach for Angra 2 and 3 Project. (author)

  14. Quality Indicators in Laboratory Medicine: from theory to practice. Preliminary data from the IFCC Working Group Project "Laboratory Errors and Patient Safety".

    Science.gov (United States)

    Sciacovelli, Laura; O'Kane, Maurice; Skaik, Younis Abdelwahab; Caciagli, Patrizio; Pellegrini, Cristina; Da Rin, Giorgio; Ivanov, Agnes; Ghys, Timothy; Plebani, Mario

    2011-05-01

    The adoption of Quality Indicators (QIs) has prompted the development of tools to measure and evaluate the quality and effectiveness of laboratory testing, first in the hospital setting and subsequently in ambulatory and other care settings. While Laboratory Medicine has an important role in the delivery of high-quality care, no consensus exists as yet on the use of QIs focussing on all steps of the laboratory total testing process (TTP), and further research in this area is required. In order to reduce errors in laboratory testing, the IFCC Working Group on "Laboratory Errors and Patient Safety" (WG-LEPS) developed a series of Quality Indicators, specifically designed for clinical laboratories. In the first phase of the project, specific QIs for key processes of the TTP were identified, including all the pre-, intra- and post-analytic steps. The overall aim of the project is to create a common reporting system for clinical laboratories based on standardized data collection, and to define state-of-the-art and Quality Specifications (QSs) for each QI independent of: a) the size of organization and type of activities; b) the complexity of processes undertaken; and c) different degree of knowledge and ability of the staff. The aim of the present paper is to report the results collected from participating laboratories from February 2008 to December 2009 and to identify preliminary QSs. The results demonstrate that a Model of Quality Indicators managed as an External Quality Assurance Program can serve as a tool to monitor and control the pre-, intra- and post-analytical activities. It might also allow clinical laboratories to identify risks that lead to errors resulting in patient harm: identification and design of practices that eliminate medical errors; the sharing of information and education of clinical and laboratory teams on practices that reduce or prevent errors; the monitoring and evaluation of improvement activities.

  15. Preliminary study of the safety and efficacy of medium-chain triglycerides for use as an intraocular tamponading agent in minipigs.

    Science.gov (United States)

    Soler, Vincent J; Laurent, Camille; Sakr, Frédéric; Regnier, Alain; Tricoire, Cyrielle; Cases, Olivier; Kozyraki, Renata; Douet, Jean-Yves; Pagot-Mathis, Véronique

    2017-08-01

    To date, only silicone oils and gases have the appropriate characteristics for use in vitreo-retinal surgery as vitreous substitutes with intraocular tamponading properties. This preliminary study evaluated the safety and efficacy of medium-chain triglycerides (MCTs) for use as a tamponading agent in minipigs. In 15 minipigs, 15 right eyes underwent vitrectomies followed by injection of MCT tamponade (day 1). Two groups were defined. In Group A (ten eyes), the surgical procedure before MCT injection included induced rhegmatogenous retinal detachment (RRD), retina flattening, and retinopexy. In Group B (five eyes), MCT was injected without inducing RRD; in these eyes, MCT was removed on day 90. Pigs were sacrificed on day 45 (Group A) or 120 (Group B). Eyes were examined on days 1, 5, 15, and 45 in both groups and on days 90 and 120 in Group B. In Group B only, we performed bilateral electroretinography examinations on days 1 and 120, and histological examinations of MCTs and controlateral eyes were performed after sacrifice. In Group A eyes (n = 9; one eye was non-assessable), on day 45, the retina was flat in seven eyes and two RRD detachments were observed in insufficiently MCT-filled eyes. In Group B, electroretinography showed no significant differences between MCT eyes and controls on days 1 or 120. Histological analyses revealed no signs of retinal toxicity. This study showed that MCT tamponade seems to be effective and safe; however, additional studies are needed before it becomes commonly used as a tamponading agent in humans.

  16. Regulatory guidance document

    International Nuclear Information System (INIS)

    1994-05-01

    The Office of Civilian Radioactive Waste Management (OCRWM) Program Management System Manual requires preparation of the OCRWM Regulatory Guidance Document (RGD) that addresses licensing, environmental compliance, and safety and health compliance. The document provides: regulatory compliance policy; guidance to OCRWM organizational elements to ensure a consistent approach when complying with regulatory requirements; strategies to achieve policy objectives; organizational responsibilities for regulatory compliance; guidance with regard to Program compliance oversight; and guidance on the contents of a project-level Regulatory Compliance Plan. The scope of the RGD includes site suitability evaluation, licensing, environmental compliance, and safety and health compliance, in accordance with the direction provided by Section 4.6.3 of the PMS Manual. Site suitability evaluation and regulatory compliance during site characterization are significant activities, particularly with regard to the YW MSA. OCRWM's evaluation of whether the Yucca Mountain site is suitable for repository development must precede its submittal of a license application to the Nuclear Regulatory Commission (NRC). Accordingly, site suitability evaluation is discussed in Chapter 4, and the general statements of policy regarding site suitability evaluation are discussed in Section 2.1. Although much of the data and analyses may initially be similar, the licensing process is discussed separately in Chapter 5. Environmental compliance is discussed in Chapter 6. Safety and Health compliance is discussed in Chapter 7

  17. CMS DOCUMENTATION

    CERN Multimedia

    CMS TALKS AT MAJOR MEETINGS The agenda and talks from major CMS meetings can now be electronically accessed from the ICMS Web site. The following items can be found on: http://cms.cern.ch/iCMS Management – CMS Weeks (Collaboration Meetings), CMS Weeks Agendas The talks presented at the Plenary Sessions. Management – CB – MB – FB Agendas and minutes are accessible to CMS members through Indico. LHCC The talks presented at the ‘CMS Meetings with LHCC Referees’ are available on request from the PM or MB Country Representative. Annual Reviews The talks presented at the 2008 Annual Reviews are posted in Indico. CMS DOCUMENTS It is considered useful to establish information on the first employment of CMS doctoral student upon completion of their theses.  Therefore it is requested that Ph.D students inform the CMS Secretariat about the nature of employment and name of their first employer. The Notes, Conference Reports and Theses published si...

  18. Repository Subsurface Preliminary Fire Hazard Analysis

    International Nuclear Information System (INIS)

    Logan, Richard C.

    2001-01-01

    This fire hazard analysis identifies preliminary design and operations features, fire, and explosion hazards, and provides a reasonable basis to establish the design requirements of fire protection systems during development and emplacement phases of the subsurface repository. This document follows the Technical Work Plan (TWP) (CRWMS M and O 2001c) which was prepared in accordance with AP-2.21Q, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities''; Attachment 4 of AP-ESH-008, ''Hazards Analysis System''; and AP-3.11Q, ''Technical Reports''. The objective of this report is to establish the requirements that provide for facility nuclear safety and a proper level of personnel safety and property protection from the effects of fire and the adverse effects of fire-extinguishing agents

  19. Omega documentation

    Energy Technology Data Exchange (ETDEWEB)

    Howerton, R.J.; Dye, R.E.; Giles, P.C.; Kimlinger, J.R.; Perkins, S.T.; Plechaty, E.F.

    1983-08-01

    OMEGA is a CRAY I computer program that controls nine codes used by LLNL Physical Data Group for: 1) updating the libraries of evaluated data maintained by the group (UPDATE); 2) calculating average values of energy deposited in secondary particles and residual nuclei (ENDEP); 3) checking the libraries for internal consistency, especially for energy conservation (GAMCHK); 4) producing listings, indexes and plots of the library data (UTILITY); 5) producing calculational constants such as group averaged cross sections and transfer matrices for diffusion and Sn transport codes (CLYDE); 6) producing and updating standard files of the calculational constants used by LLNL Sn and diffusion transport codes (NDFL); 7) producing calculational constants for Monte Carlo transport codes that use group-averaged cross sections and continuous energy for particles (CTART); 8) producing and updating standard files used by the LLNL Monte Carlo transport codes (TRTL); and 9) producing standard files used by the LANL pointwise Monte Carlo transport code MCNP (MCPOINT). The first four of these functions and codes deal with the libraries of evaluated data and the last five with various aspects of producing calculational constants for use by transport codes. In 1970 a series, called PD memos, of internal and informal memoranda was begun. These were intended to be circulated among the group for comment and then to provide documentation for later reference whenever questions arose about the subject matter of the memos. They have served this purpose and now will be drawn upon as source material for this more comprehensive report that deals with most of the matters covered in those memos.

  20. Omega documentation

    International Nuclear Information System (INIS)

    Howerton, R.J.; Dye, R.E.; Giles, P.C.; Kimlinger, J.R.; Perkins, S.T.; Plechaty, E.F.

    1983-08-01

    OMEGA is a CRAY I computer program that controls nine codes used by LLNL Physical Data Group for: 1) updating the libraries of evaluated data maintained by the group (UPDATE); 2) calculating average values of energy deposited in secondary particles and residual nuclei (ENDEP); 3) checking the libraries for internal consistency, especially for energy conservation (GAMCHK); 4) producing listings, indexes and plots of the library data (UTILITY); 5) producing calculational constants such as group averaged cross sections and transfer matrices for diffusion and Sn transport codes (CLYDE); 6) producing and updating standard files of the calculational constants used by LLNL Sn and diffusion transport codes (NDFL); 7) producing calculational constants for Monte Carlo transport codes that use group-averaged cross sections and continuous energy for particles (CTART); 8) producing and updating standard files used by the LLNL Monte Carlo transport codes (TRTL); and 9) producing standard files used by the LANL pointwise Monte Carlo transport code MCNP (MCPOINT). The first four of these functions and codes deal with the libraries of evaluated data and the last five with various aspects of producing calculational constants for use by transport codes. In 1970 a series, called PD memos, of internal and informal memoranda was begun. These were intended to be circulated among the group for comment and then to provide documentation for later reference whenever questions arose about the subject matter of the memos. They have served this purpose and now will be drawn upon as source material for this more comprehensive report that deals with most of the matters covered in those memos

  1. Development of a procedure for qualitative and quantitative evaluation of human factors as a part of probabilistic safety assessments of nuclear power plants. Part B. Technical documentation

    International Nuclear Information System (INIS)

    Richei, A.

    1998-01-01

    As international studies have shown, accidents in plants are increasingly caused by combinations of technical failures and human errors. Therefore careful investigations of man-machine-interactions to determine human reliability are gaining importance worldwide. Regarding nuclear power plants such investigations are usually carried out within the scope of probabilistic safety assessments. A great number of procedures to evaluate human factors has been developed up to now. However, none of them is able to take into account the whole spectrum of requirements - as for instance transferability of date to other plants, analysis of weak points, and evaluation of cognitive tasks - for a complete and reliable probabilistic safety assessment. Based on an advanced model for a man-machine-system, the Human Error Rate Assessment and Optimizing System (HEROS) and a corresponding expert system of the same name are introduced. This expert system enables the quantification of human error probabilities for plant operator actions on the one hand and is also capable of providing quantitative statements regarding the optimization of man-machine-system in terms of human error probability minimization on the other one. Three relevant evaluation levels, i.e. 'Management Structure', 'Working Environment' and 'Man-Machine-Interface', are derived from a model of the man-machine-system. Linguistic variables are assigned to all performance shaping factors at these levels. These variables are used to establish a rule-based expert system. The knowledge bases of this system are represented by rules. Processing of these rules is carried out by means of the fuzzy set theory, after provision of relevant data for a particular personal action to be evaluated. This procedure enables a simple and effective use of ergonomic studies as the relevant database, which is also transferable to other plants with any design. The expert system consist in total of 16 rule bases in which all ascertainable and

  2. Hanford surplus facilities hazards identification document

    International Nuclear Information System (INIS)

    Egge, R.G.

    1997-01-01

    This document provides general safety information needed by personnel who enter and work in surplus facilities managed by Bechtel Hanford, Inc. The purpose of the document is to enhance access control of surplus facilities, educate personnel on the potential hazards associated with these facilities prior to entry, and ensure that safety precautions are taken while in the facility

  3. SAFETY INSTRUCTION AND SAFETY NOTE

    CERN Multimedia

    TIS Secretariat

    2002-01-01

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

  4. Preliminary site description Laxemar stage 2.1. Feedback for completion of the site investigation including input from safety assessment and repository engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-09-15

    The Laxemar subarea is the focus for the complete site investigations in the Simpevarp area. The south and southwestern parts of the subarea (the so-called 'focused area') have been designated for focused studies during the remainder of the site investigations. This area, some 5.3 square kilometres in size, is characterised on the surface by an arc shaped body of quartz monzodiorite gently dipping to the north, flanked in the north and south by Aevroe granite. The current report documents work conducted during stage 2.1 of the site-descriptive modelling of the Laxemar subarea. The primary objective of the work performed is to provide feedback to the site investigations at Laxemar to ensure that adequate and timely data and information are obtained during the remaining investigation stage. The work has been conducted in cooperation with the site investigation team at Laxemar and representatives from safety assessment and repository engineering. The principal aim of this joint effort has been to safeguard that adequate data are collected that resolve the remaining issues/uncertainties which are of importance for repository layout and long-term safety. The proposed additional works presented in this report should be regarded as recommended additions and/or modifications in relation to the CSI programme published early 2006. The overall conclusion of the discipline-wise review of critical issues is that the CSI programme overall satisfies the demands to resolve the remaining uncertainties. This is interpreted to be partly a result of the close interaction between the site modelling team, site investigation team and the repository engineering teams, which has been in operation since early 2005. In summary, the performed interpretations and modelling have overall confirmed the version 1.2 results. The exception being Hydrogeology where the new Laxemar 2.1 borehole data suggest more favourable conditions in the south and west parts of the focused area compared

  5. Preliminary site description Laxemar stage 2.1. Feedback for completion of the site investigation including input from safety assessment and repository engineering

    International Nuclear Information System (INIS)

    2006-09-01

    The Laxemar subarea is the focus for the complete site investigations in the Simpevarp area. The south and southwestern parts of the subarea (the so-called 'focused area') have been designated for focused studies during the remainder of the site investigations. This area, some 5.3 square kilometres in size, is characterised on the surface by an arc shaped body of quartz monzodiorite gently dipping to the north, flanked in the north and south by Aevroe granite. The current report documents work conducted during stage 2.1 of the site-descriptive modelling of the Laxemar subarea. The primary objective of the work performed is to provide feedback to the site investigations at Laxemar to ensure that adequate and timely data and information are obtained during the remaining investigation stage. The work has been conducted in cooperation with the site investigation team at Laxemar and representatives from safety assessment and repository engineering. The principal aim of this joint effort has been to safeguard that adequate data are collected that resolve the remaining issues/uncertainties which are of importance for repository layout and long-term safety. The proposed additional works presented in this report should be regarded as recommended additions and/or modifications in relation to the CSI programme published early 2006. The overall conclusion of the discipline-wise review of critical issues is that the CSI programme overall satisfies the demands to resolve the remaining uncertainties. This is interpreted to be partly a result of the close interaction between the site modelling team, site investigation team and the repository engineering teams, which has been in operation since early 2005. In summary, the performed interpretations and modelling have overall confirmed the version 1.2 results. The exception being Hydrogeology where the new Laxemar 2.1 borehole data suggest more favourable conditions in the south and west parts of the focused area compared with the

  6. Deep repository for long-lived low- and intermediate-level waste in Sweden (SFL 3-5): An international peer review of SKB 's preliminary safety assessment

    International Nuclear Information System (INIS)

    Chapman, N.; Apted, M.; Glasser, F.; Voss, C.

    2000-10-01

    The SKB safety assessment of the SFL 3-5 repository (the planned deep repository for long-lived low- and intermediate level waste) can be read in two contexts: as a preliminary evaluation of the performance and design options for a repository that will not be required for perhaps forty years; or as an evaluation of a repository that might need to be sited together with the SFL 2 spent fuel repository, and whose nature and performance might thus need to be understood to a level that can be used to make wider programmatic decisions during the next five years. These two 'assessment contexts' are quite different, and an overarching issue is the fact that it was not clear to the review team which view to take. Apparently, SKB would tend towards the first context. However, it is not at all apparent to the reviewers why the second context should not be the predominant driver in the near future. The review team notes that the SFL 3-5 repository, as modelled by SKB, gives rise to potentially perceptible radionuclide releases to the environment on a timescale of hundreds of years after closure. This is in contrast to the SR 97 assessment for the SFL 2 spent fuel repository, which base scenario predicts no releases over a million year timescale. It is clear that according to SKB's SR97 and SFL3-5 analyses, for co-located facilities, it is this repository that has the potential for real radiological impacts in the immediate future. An initial recommendation from the review, is that SKB and the regulatory authorities consider which context is appropriate to the current status of the Swedish programme. This is important, because an overall impression of the reviewers is that the analysis would not be 'fit for purpose' if it were needed to assist with decision-making by SKB or the regulatory agencies. There are too many unanswered questions, and the overall impression of the safety concept is one of some fragility. Because there is no real design basis presented, no thorough

  7. Technical basis document for external events

    International Nuclear Information System (INIS)

    OBERG, B.D.

    2003-01-01

    This document supports the Tank Farms Documented Safety Analysis and presents the technical basis for the FR-equencies of externally initiated accidents. The consequences of externally initiated events are discussed in other documents that correspond to the accident that was caused by the external event. The external events include aircraft crash, vehicle accident, range fire, and rail accident

  8. Safety; Avertissement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  9. Melter Disposal Strategic Planning Document

    Energy Technology Data Exchange (ETDEWEB)

    BURBANK, D.A.

    2000-09-25

    This document describes the proposed strategy for disposal of spent and failed melters from the tank waste treatment plant to be built by the Office of River Protection at the Hanford site in Washington. It describes program management activities, disposal and transportation systems, leachate management, permitting, and safety authorization basis approvals needed to execute the strategy.

  10. RHF RELAP5 model and preliminary loss-of-offsite-power simulation results for LEU conversion

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Bergeron, A. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Dionne, B. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Thomas, F. [Institut Laue-Langevin (ILL), Grenoble (Switzerland). RHF Reactor Dept.

    2014-08-01

    The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

  11. General safety considerations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-09-01

    This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

  12. General safety considerations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

  13. General safety considerations

    International Nuclear Information System (INIS)

    2001-01-01

    This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

  14. General safety considerations

    International Nuclear Information System (INIS)

    1998-01-01

    This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

  15. Technical basis document for natural event hazards

    International Nuclear Information System (INIS)

    CARSON, D.M.

    2003-01-01

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for natural event hazards (NEH)-initiated representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

  16. Planned activities to improve safety

    International Nuclear Information System (INIS)

    1998-01-01

    This document presents the fulfilling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 6 of the document contains some details about the planed activities to safety improvements

  17. L-Band Digital Aeronautical Communications System Engineering - Initial Safety and Security Risk Assessment and Mitigation

    Science.gov (United States)

    Zelkin, Natalie; Henriksen, Stephen

    2011-01-01

    This document is being provided as part of ITT's NASA Glenn Research Center Aerospace Communication Systems Technical Support (ACSTS) contract NNC05CA85C, Task 7: "New ATM Requirements--Future Communications, C-Band and L-Band Communications Standard Development." ITT has completed a safety hazard analysis providing a preliminary safety assessment for the proposed L-band (960 to 1164 MHz) terrestrial en route communications system. The assessment was performed following the guidelines outlined in the Federal Aviation Administration Safety Risk Management Guidance for System Acquisitions document. The safety analysis did not identify any hazards with an unacceptable risk, though a number of hazards with a medium risk were documented. This effort represents a preliminary safety hazard analysis and notes the triggers for risk reassessment. A detailed safety hazards analysis is recommended as a follow-on activity to assess particular components of the L-band communication system after the technology is chosen and system rollout timing is determined. The security risk analysis resulted in identifying main security threats to the proposed system as well as noting additional threats recommended for a future security analysis conducted at a later stage in the system development process. The document discusses various security controls, including those suggested in the COCR Version 2.0.

  18. Areva - 2016 Reference document

    International Nuclear Information System (INIS)

    2017-01-01

    Areva supplies high added-value products and services to support the operation of the global nuclear fleet. The company is present throughout the entire nuclear cycle, from uranium mining to used fuel recycling, including nuclear reactor design and operating services. Areva is recognized by utilities around the world for its expertise, its skills in cutting-edge technologies and its dedication to the highest level of safety. Areva's 36,000 employees are helping build tomorrow's energy model: supplying ever safer, cleaner and more economical energy to the greatest number of people. This Reference Document contains information on Areva's objectives, prospects and development strategies. It contains estimates of the markets, market shares and competitive position of Areva

  19. Documents and legal texts

    International Nuclear Information System (INIS)

    2013-01-01

    This section reprints a selection of recently published legislative texts and documents: - Russian Federation: Federal Law No.170 of 21 November 1995 on the use of atomic energy, Adopted by the State Duma on 20 October 1995; - Uruguay: Law No.19.056 On the Radiological Protection and Safety of Persons, Property and the Environment (4 January 2013); - Japan: Third Supplement to Interim Guidelines on Determination of the Scope of Nuclear Damage resulting from the Accident at the Tokyo Electric Power Company Fukushima Daiichi and Daini Nuclear Power Plants (concerning Damages related to Rumour-Related Damage in the Agriculture, Forestry, Fishery and Food Industries), 30 January 2013; - France and the United States: Joint Statement on Liability for Nuclear Damage (Aug 2013); - Franco-Russian Nuclear Power Declaration (1 November 2013)

  20. ESCMID Study Group for Infections in Compromised Hosts (ESGICH) Consensus Document on the safety of targeted and biological therapies: an infectious diseases perspective (Intracellular signaling pathways: tyrosine kinase and mTOR inhibitors).

    Science.gov (United States)

    Reinwald, M; Silva, J T; Mueller, N J; Fortún, J; Garzoni, C; de Fijter, J W; Fernández-Ruiz, M; Grossi, P; Aguado, J M

    2018-06-01

    The present review is part of the European Society of Clinical Microbiology and Infectious Diseases (ESCMID) Study Group for Infections in Compromised Hosts (ESGICH) Consensus Document on the safety of targeted and biologic therapies. To review, from an infectious diseases perspective, the safety profile of therapies targeting different intracellular signaling pathways and to suggest preventive recommendations. Computer-based Medline searches with MeSH terms pertaining to each agent or therapeutic family. Although BCR-ABL tyrosine kinase inhibitors modestly increase the overall risk of infection, dasatinib has been associated with cytomegalovirus and hepatitis B virus reactivation. BRAF/MEK kinase inhibitors do not significantly affect infection susceptibility. The effect of Bruton tyrosine kinase inhibitors (ibrutinib) among patients with B-cell malignancies is difficult to distinguish from that of previous immunosuppression. However, cases of Pneumocystis jirovecii pneumonia (PCP), invasive fungal infection and progressive multifocal leukoencephalopathy have been occasionally reported. Because phosphatidylinositol-3-kinase inhibitors (idelalisib) may predispose to opportunistic infections, anti-Pneumocystis prophylaxis and prevention strategies for cytomegalovirus are recommended. No increased rates of infection have been observed with venetoclax (antiapoptotic protein Bcl-2 inhibitor). Therapy with Janus kinase inhibitors markedly increases the incidence of infection. Pretreatment screening for chronic hepatitis B virus and latent tuberculosis infection must be performed, and anti-Pneumocystis prophylaxis should be considered for patients with additional risk factors. Cancer patients receiving mTOR inhibitors face an increased incidence of overall infection, especially those with additional risk factors (prior therapies or delayed wound healing). Specific preventive approaches are warranted in view of the increased risk of infection associated with some of the

  1. An analysis of electronic document management in oncology care.

    Science.gov (United States)

    Poulter, Thomas; Gannon, Brian; Bath, Peter A

    2012-06-01

    In this research in progress, a reference model for the use of electronic patient record (EPR) systems in oncology is described. The model, termed CICERO, comprises technical and functional components, and emphasises usability, clinical safety and user acceptance. One of the functional components of the model-an electronic document and records management (EDRM) system-is monitored in the course of its deployment at a leading oncology centre in the UK. Specifically, the user requirements and design of the EDRM solution are described.The study is interpretative and forms part a wider research programme to define and validate the CICERO model. Preliminary conclusions confirm the importance of a socio-technical perspective in Onco-EPR system design.

  2. ITER Safety and Licensing

    International Nuclear Information System (INIS)

    Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

    2006-01-01

    The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

  3. Traceability of Software Safety Requirements in Legacy Safety Critical Systems

    Science.gov (United States)

    Hill, Janice L.

    2007-01-01

    How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

  4. Creation, organizing and development of the French nuclear documentation network

    International Nuclear Information System (INIS)

    Vergnes, Gisele; Cheron, Christiane; Guilloux, Raymond

    1974-01-01

    This historical account covers the creation of the French nuclear documentation network, the preliminary research, aims, structures and first accomplishments of the Association Francaise de Documentation et d'Information Nucleaire (AFDIN) (French Association of Nuclear Documentation and Information) and its relations with the international system INIS [fr

  5. A preliminary study on the application of system dynamics methodology to organizational safety in nuclear power plants: Learning from past models

    Energy Technology Data Exchange (ETDEWEB)

    Do, Giang [Sol Bridge International School of Business, Daejeon (Korea, Republic of); Kim, Sakil; Lee, Yong Hee; Lee, Yong Hee [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Besides technical design, organizational and human factor are of increasing interest in literature on nuclear safety. Among the methodologies employed to study these factors, System Dynamics (SD) is considered to be suitable for addressing the complexity and dynamicity of the organizational system in nuclear power plants (NPPs). In the following sections, the method will be described and its several prior applications to studying organizational safety will be introduced. An SD model with emphasis on the role of organizational learning in organizational safety will be presented.

  6. Exploratory shaft facility preliminary designs - Permian Basin

    International Nuclear Information System (INIS)

    1983-09-01

    The purpose of the Preliminary Design Report, Permian Basin, is to provide a description of the preliminary design for an Exploratory Shaft Facility in the Permian Basin, Texas. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers are included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description, and Construction Cost Estimate. 30 references, 13 tables

  7. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    KOPELIC, S.D.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  8. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  9. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.

    1999-01-01

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  10. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.; PIEPHO, M.G.

    2000-01-01

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  11. Gas-cooled breeder reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

    1981-01-15

    The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

  12. Review of the Federal Motor Carrier Safety Regulations for Automated Commercial Vehicles: Preliminary Assessment of Interpretation and Enforcement Challenges, Questions, and Gaps

    Science.gov (United States)

    2018-03-01

    The Volpe National Transportation Systems Center (Volpe) reviewed the Federal Motor Carrier Safety Regulations (FMCSRs) to identify compliance and enforcement challenges related to the operation of automated commercial vehicles (CMVs) in interstate c...

  13. Preliminary hazard classification for Building 107-N

    International Nuclear Information System (INIS)

    Kloster, G.L.; Smith, R.L.

    1997-06-01

    Deactivation activities are planned for Building 107-N (Basin Recirculation Building). This document establishes the preliminary hazard classification (PHC) for the 100-N Area facility segment that includes this building.To establish the PHC, the inventories of radioactive and nonradioactive hazardous materials present within Building 107-N are identified and then compared to the corresponding threshold quantity values in DOE (1992) and reportable quantity values in 40 CFR 302.4. In this evaluation, no credit is taken for the form, location, and dispersibility of the materials; for their interaction with available energy sources; or for safety features that could prevent or mitigate a radioactive release. The result of this effort concluded that the PHC for Building 107-N is Nuclear Category 3

  14. Interim process report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Sellin, Patrick

    2004-08-01

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

  15. Interim process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Sellin, Patrick (ed.)

    2004-08-01

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

  16. Safety review and approval process for the TFTR

    International Nuclear Information System (INIS)

    Levine, J.D.; Howe, H.J.; Howe, K.E.

    1983-01-01

    The design, construction, and operation of the Tokamak Fusion Test Reactor (TFTR) has undergone an extensive safety and enviromental analysis involving Princeton Plasma Physics Laboratory (PPPL), the U.S. Department of Energy (DOE), the Ebasco/Grumman Industrial Subcontractor Team, and other organizations. This analysis, which is continuing during the TFTR operational phase, has been facilitated by the preparation, review and approval of several documents, including an Environmental Statement (Draft and Final), a Preliminary Safety Analysis Report (PSAR), a Final Safety Analysis Report (FSAR), Operations Safety Requirements (OSRs) and Safety Requirements (SRs), and various Operating and Maintenance Manuals. Through TFTR Safety Group participation in formal system design evaluations, change control boards, and reviews of project procurement and installation documentation, the TFTR Management Configuration Control System assures that all aspects of the project, including proposed design, installation and operational changes, receive prompt and thorough safety analyses. These efforts will continue as the TFTR Program moves into the neutral beam and D-T operational phases. The safety review and approval experience that has been acquired on the TFTR Project should serve as a foundation for similar efforts on future fusion devices

  17. International Standardization of Library and Documentation Techniques.

    Science.gov (United States)

    International Federation for Documentation, The Hague (Netherlands).

    This comparative study of the national and international standards, rules and regulations on library and documentation techniques adopted in various countries was conducted as a preliminary step in determining the minimal bases for facilitating national and international cooperation between documentalists and librarians. The study compares and…

  18. Gaia DR1 documentation

    Science.gov (United States)

    van Leeuwen, F.; de Bruijne, J. H. J.; Arenou, F.; Comoretto, G.; Eyer, L.; Farras Casas, M.; Hambly, N.; Hobbs, D.; Salgado, J.; Utrilla Molina, E.; Vogt, S.; van Leeuwen, M.; Abreu, A.; Altmann, M.; Andrei, A.; Babusiaux, C.; Bastian, U.; Biermann, M.; Blanco-Cuaresma, S.; Bombrun, A.; Borrachero, R.; Brown, A. G. A.; Busonero, D.; Busso, G.; Butkevich, A.; Cantat-Gaudin, T.; Carrasco, J. M.; Castañeda, J.; Charnas, J.; Cheek, N.; Clementini, G.; Crowley, C.; Cuypers, J.; Davidson, M.; De Angeli, F.; De Ridder, J.; Evans, D.; Fabricius, C.; Findeisen, K.; Fleitas, J. M.; Gracia, G.; Guerra, R.; Guy, L.; Helmi, A.; Hernandez, J.; Holl, B.; Hutton, A.; Klioner, S.; Lammers, U.; Lecoeur-Taïbi, I.; Lindegren, L.; Luri, X.; Marinoni, S.; Marrese, P.; Messineo, R.; Michalik, D.; Mignard, F.; Montegriffo, P.; Mora, A.; Mowlavi, N.; Nienartowicz, K.; Pancino, E.; Panem, C.; Portell, J.; Rimoldini, L.; Riva, A.; Robin, A.; Siddiqui, H.; Smart, R.; Sordo, R.; Soria, S.; Turon, C.; Vallenari, A.; Voss, H.

    2017-12-01

    94000 Hipparcos stars in the primary data set, the proper motion standard errors are much smaller, at about 0.06 mas yr^-1. For the secondary astrometric data set, the typical standard error on the positions is 10 mas. The median standard errors on the mean G-band magnitudes range from the milli-magnitude level to 0.03 mag over the magnitude range 5 to 20.7. The DPAC undertook an extensive validation of Gaia DR1 which confirmed that this data release represents a major advance in the mapping of the skies and the availability of basic stellar data that form the foundation of observational astrophysics. However, as a consequence of the very preliminary nature of this first Gaia data release, there are a number of important limitations to the data quality. These limitations are documented in the Astronomy & Astrophysics papers that accompany Gaia DR1, with further information provided in this documentation. The reader is strongly encouraged to read about these limitations and to carefully consider them before drawing conclusions from the data. This Gaia DR1 documentation complements the peer-reviewed papers that accompany the release in a Special Issue of Astronomy & Astrophysics. The papers form the primary documentation for the data release and they are frequently referenced throughout the text.

  19. 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.

  20. Areva - 2014 Reference document

    International Nuclear Information System (INIS)

    2015-01-01

    Areva supplies high added-value products and services to support the operation of the global nuclear fleet. The company is present throughout the entire nuclear cycle, from uranium mining to used fuel recycling, including nuclear reactor design and operating services. Areva is recognized by utilities around the world for its expertise, its skills in cutting-edge technologies and its dedication to the highest level of safety. Areva's 44,000 employees are helping build tomorrow's energy model: supplying ever safer, cleaner and more economical energy to the greatest number of people. This Reference Document contains information on Areva's objectives, prospects and development strategies. It contains estimates of the markets, market shares and competitive position of Areva. Contents: 1 - Person responsible; 2 - Statutory auditors; 3 - Selected financial information; 4 - Risk factors; 5 - Information about the issuer; 6 - Business overview; 7 - Organizational structure; 8 - Property, plant and equipment; 9 - Analysis of and comments on the group's financial position and performance; 10 - Capital resources; 11 - Research and development programs, patents and licenses; 12 - Trend information; 13 - Profit forecasts; 14 - Administrative, management and supervisory bodies and senior management; 15 - Compensation and benefits; 16 - Functioning of administrative, management and supervisory bodies and senior management; 17 - Employees; 18 - Principal shareholders; 19 - Transactions with related parties; 20 - Financial information concerning assets, financial positions and financial performance; 21 - Additional information; 22 - Major contracts; 23 - Third party information, statements by experts and declarations of interest; 24 - Documents on display; 25 - information on holdings; appendix: Report of the Chairman of the Board of Directors on governance, internal control procedures and risk management, Statutory Auditors' report, Corporate social

  1. Technical documentation challenges in aviation maintenance : a proceedings report.

    Science.gov (United States)

    2012-11-01

    The 2012 Technical Documentation workshop addressed both problems and solutions associated with technical : documentation for maintenance. These issues are known to cause errors, rework, maintenance delays, other : safety hazards, and FAA administrat...

  2. Fluor Daniel Hanford contract standards/requirements identification document

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, G.L.

    1997-04-24

    This document, the Standards/Requirements Identification Document (S/RID) for the Fluor Daniel Hanford Contract, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment.

  3. Fluor Daniel Hanford company standards requirements identification document

    International Nuclear Information System (INIS)

    Bennett, G.L.

    1997-01-01

    This document, the Standards/Requirements Identification Document (S/RID) for the Fluor Daniel Hanford Contract, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment

  4. Preliminary hazards analysis of thermal scrap stabilization system. Revision 1

    International Nuclear Information System (INIS)

    Lewis, W.S.

    1994-01-01

    This preliminary analysis examined the HA-21I glovebox and its supporting systems for potential process hazards. Upon further analysis, the thermal stabilization system has been installed in gloveboxes HC-21A and HC-21C. The use of HC-21C and HC-21A simplified the initial safety analysis. In addition, these gloveboxes were cleaner and required less modification for operation than glovebox HA-21I. While this document refers to glovebox HA-21I for the hazards analysis performed, glovebox HC-21C is sufficiently similar that the following analysis is also valid for HC-21C. This hazards analysis document is being re-released as revision 1 to include the updated flowsheet document (Appendix C) and the updated design basis (Appendix D). The revised Process Flow Schematic has also been included (Appendix E). This Current revision incorporates the recommendations provided from the original hazards analysis as well. The System Design Description (SDD) has also been appended (Appendix H) to document the bases for Safety Classification of thermal stabilization equipment

  5. FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS

    Energy Technology Data Exchange (ETDEWEB)

    GARVIN, L. J.; JENSEN, M. A.

    2004-04-13

    This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

  6. Preliminary results from a field experiment on e-bike safety : Speed choice and mental workload for middle-aged and elderly cyclists

    NARCIS (Netherlands)

    Twisk, D.A.M.; Boele, M.J.; Vlakveld, W.P.; Christoph, M.; Sikkema, R.; Remij, R.; Schwab, A.L.

    2013-01-01

    To study the safety of e-bikes for the elderly, an experimental field study was conducted, using instrumented bicycles and comparing two age groups: older cyclists, n= 29, mean age = 70, SD = 4.2 and middle-aged cyclists, n = 29, mean age = 38, SD = 4.3. All were regular cyclists. They rode a fixed

  7. Preliminary Hazard Classification of the 1714-N, Lead Storage

    International Nuclear Information System (INIS)

    Kerr, N. R.

    1999-01-01

    The 1714-N, -NA and -NB is a building segment that was deactivated under the N Area Deactivation Project. During the deactivation the building was designated as an area to store recycled or reused lead products. This document presents the Preliminary Hazard Classification (PHC) for the continued storage of lead products by Bechtel Hanford, Inc. (BHI). Two types of hazardous substances are the focus of this PHC: lead and residual radiological contamination. An evaluation contained in this PHC concludes that there is little risk from the remaining hazardous substances. It was further concluded that standard institutional controls that are implemented under the BHI contract provide adequate protection to people and the environment. No further safety analysis documentation is required for the continued lead storage

  8. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

  9. 76 FR 5537 - Electronic On-Board Recorders and Hours of Service Supporting Documents

    Science.gov (United States)

    2011-02-01

    ... supporting documents influence driver operational safety rather than vehicular and mechanical safety... generate greater traffic congestion than a crash involving a CMV transporting other cargoes. Although these...

  10. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  11. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  12. Geochemical analysis of the sealing system. Technical Report to work package 9.1.2. Preliminary safety case of the Gorleben site (VSG)

    Energy Technology Data Exchange (ETDEWEB)

    Xie, Mingliang; Herbert, Horst-Juergen

    2012-02-15

    The geochemical reaction simulation of ground water and brines with sealing materials for a proposed repository at the Gorleben site is a task under the framework of the VSG project. The calculations presented in this report are aimed to provide a preliminary evaluation of the geochemical stability of three potential engineering barrier materials foreseen in the shaft sealing system in case of groundwater and brine intrusion. The long-term stability of these materials is a key issue for the sealing function of the shaft seals. This is governed by many factors such as geotechnical, hydraulic and geochemical processes. In order to better understand the potential effect of geochemical processes on the long-term properties of these sealing materials, geochemical simulations of the potential interactions between groundwater and brine and shaft sealing materials were performed.

  13. Efficacy and safety of escitalopram versus desvenlafaxine in the treatment of major depression: A preliminary 1-year prospective randomized open label comparative trial

    Directory of Open Access Journals (Sweden)

    Brij Mohan Gupta

    2016-01-01

    Full Text Available Aim and Objective: To compare efficacy and safety of escitalopram with desvenlafaxine in the treatment of major depression.Materials and Methods: A total of 60 patients of depression were randomized into two groups after meeting inclusion criterion. In the first 3 weeks, escitalopram 10 mg/day was given and then 20 mg/day for the next 3 weeks in group 1 (n = 30. Desvenlafaxine in the first 3 weeks was given 50 mg/day and 100 mg/day for the next 3 weeks in group 2 (n = 30. The parameters evaluated during the study were efficacy assessments byHamilton Scale of Rating Depression (HAM-D, Hamilton Rating Scale of Anxiety (HAM-A, and Clinical Global Impression (CGI. Safety assessments were done by UKU-scale. Results: Escitalopram and desvenlafaxine significantly (P < 0.001, reduced HAM-D, HAM-A, and CGI scores from their respective base lines. However, on comparison failed show any statistical difference at 3 and 6 weeks of treatment. Escitalopram and desvenlafaxine were both found to be safe and well-tolerated and there was not much difference between the two groups as evident from UKU Scale and their effect on various biochemical parameters. Conclusion: The present study demonstrated similar efficacy and safety in reducing depression and anxiety with both escitalopram and desvenlafaxine, but clinical superiority of one drug over the other cannot be concluded due to limitations of the small sample size.

  14. A Preliminary Study for Safety Shutter design to Protect Streaming of Residual Radiation Passing through Beamline in Pre-Separator Room of ISOL

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Woo; Kim, Do Hyun; Kim, Song Hyun; Shin, Chang Ho; Nam, Shin Woo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    RAON is a heavy ion accelerator under construction by the Institute for Basic Science (IBS) in Korea. As one part of the RAON accelerator, ISOL is a facility to generate and separate rare isotopes for various experiments. In ISOL facility, isotopes generated from the reaction between 70 MeV proton beam and UC{sub 2} target are transferred to pre-separator room. Almost all isotopes accumulated in slit of pre-separator except specific isobars, which are set for experiments. Residual radiations are generated from accumulated isotopes because these isotopes are unstable. Streaming of residual radiation by the beamline is weak point for radiation shielding design. In this study, safety shutter was designed. Residual radiation generated from accumulated isotopes at slit of pre-separator was estimated using following conditions: (1) the isotopes generated by proton-target reactions are accumulated at slit with 10 % accumulation rate; (2) it was assumed that the radioactive isotopes are uniformly distributed in the cylindrical slit which have 1 cm height and 15 diameter. To design optimized safety shutter, following steps were performed: (1) thickness and diameter of the bulk shield material were evaluated to optimize safety shutter material; (2) additional shielding structure was proposed using dose contribution of each additional shielding wall.

  15. Bechtel Hanford, Inc./ERC team health and safety plan Environmental Restoration Disposal Facility operations

    International Nuclear Information System (INIS)

    Turney, S.R.

    1996-02-01

    A comprehensive safety and health program is essential for reducing work-related injuries and illnesses while maintaining a safe and health work environment. This document establishes Bechtel Hanford, Inc. (BHI)/Environmental Restoration Contractor (ERC) team requirements, policies, and procedures and provides preliminary guidance to the Environmental Restoration Disposal Facility (ERDF) subcontractor for use in preparing essential safety and health documents. This health and safety plan (HASP) defines potential safety and health issues associated with operating and maintaining the ERDF. A site-specific HASP shall be developed by the ERDF subcontractor and shall be implemented before operations and maintenance work can proceed. An activity hazard analysis (AHA) shall also be developed to provide procedures to identify, assess, and control hazards or potential incidents associated with specific operations and maintenance activities

  16. Registration document 2005; Document de reference 2005

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This reference document of Gaz de France provides information and data on the Group activities in 2005: financial informations, business, activities, equipments factories and real estate, trade, capital, organization charts, employment, contracts and research programs. (A.L.B.)

  17. Preliminary results from a field experiment on e-bike safety : speed choice and mental workload for middle-aged and elderly cyclists. Paper presented at the International Cycling Safety Conference 2013, Helmond, The Netherland, 20-21 November 2013.

    NARCIS (Netherlands)

    Twisk, D.A.M. Boele, M.J. Vlakveld, W.P. Christoph, M. Sikkema, R. Remij, R. & Schwab, A.L.

    2014-01-01

    To study the safety of e-bikes for the elderly, an experimental field study was conducted, using instrumented bicycles and comparing two age groups: older cyclists, n= 29, mean age = 70, SD = 4.2 and middle-aged cyclists, n = 29, mean age = 38, SD = 4.3. All were regular cyclists. They rode a fixed

  18. 2002 reference document; Document de reference 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This 2002 reference document of the group Areva, provides information on the society. Organized in seven chapters, it presents the persons responsible for the reference document and for auditing the financial statements, information pertaining to the transaction, general information on the company and share capital, information on company operation, changes and future prospects, assets, financial position, financial performance, information on company management and executive board and supervisory board, recent developments and future prospects. (A.L.B.)

  19. Is dipyridamole test equivalent to exercise testing for the performing of myocardial tomo-scintigraphy? preliminary results of a systematic comparative analysis in patients having documented episodes of myocardial ischemia

    International Nuclear Information System (INIS)

    David, N.; Quiri, N.; Hassan, N.; Arsena, T.; Py, M.; Olivier, P.; Karcher, G.; Bertrand, A.; Houriez, P.; Grentzinger, A.; Angioi, M.; Danchin, N.; Juilliere, Y.

    1997-01-01

    Dipyridamole test is generally used rather than exercise testing when myocardial tomo-scintigraphy (MTS) is performed with patients unable to perform a maximal exercise test. However, this choice has never been validated by a systematic comparative analysis of the results provided by these two stress techniques with patients having a documented myocardial ischemia. We have included 20 patients who had a known coronary artery disease, for whom exercise test was positive at the time of an exercise MTS-TI201 performed in our department and who underwent an additional MTS-TI201 after intravenous administration of 0,56 mg/kg of dipyridamole and low-level exercise testing (40 W). The extent of perfusion abnormalities, observed after dipyvirdamole, was not significantly different from that evidenced at exercise, even in the analysis restricted to the eight patients who had a sub-maximal test (<80 % of predicted maximal heart rate) at the time of exercise-MTS (% of left ventricle: 17 ± 13 vs 15 ± 8). By contrast, there were important individual variations: a difference in defect-extent (≥ 10 % of left ventricle) was observed between the two tests for 11 patients (55 %), the largest defect being that observed at exercise in six cases, and after dipyridamole in five cases. With patients having exercise myocardial ischemia, the MTS obtained after dipyridamole are frequently very different from those performed after exercise, and the criterion of a low maximal heart rate doses not imply an underestimation of the perfusion abnormalities observed at exercise. (authors)

  20. Integrated Safety in ''SARAF'

    International Nuclear Information System (INIS)

    Dickstein, P.; Grof, Y.; Machlev, M.; Pernick, A.

    2004-01-01

    As of the very early stages of the accelerator project at the Soreq Nuclear Research Center ''SARAF'' a safety group was established which has been an inseparable participant in the planning and design of the new facility. The safety group comprises of teams responsible for the shielding, radiation protection and general industrial safety aspects of ''SARAF''. The safety group prepared and documented the safety envelope for the accelerator, dealing with the safety requirements and guidelines for the first, pre-operational, stages of the project. The safety envelope, though based upon generic principles, took into account the accelerator features and the expected modes of operation. The safety envelope was prepared in a hierarchical structure, containing Basic Principles, Basic Guidelines, General Principles for Safety Implementation, Safety Requirements and Safety Underlining Issues. The above safety envelope applies to the entire facility, which entails the accelerator itself and the experimental areas and associated plant and equipment utilizing and supporting the production of the accelerated particle beams

  1. IAEA activities to prepare safety codes and guides for thermal neutron nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1977-01-01

    In accordance with the programme presented to, and endorsed by, the eighteenth General Conference in September 1974, the IAEA is now developing a complete set of safety codes and guides that will represent recommendations for the safety of thermal neutron power plants. The safety codes outline the minimum requirements for achieving this safety, and the safety guides set forth the criteria, procedures and methods to implement the safety codes. The whole programme is directed towards the five areas of Governmental Organization, Siting, Design, Operation, and Quality Assurance. One Scientific Secretary from the Agency Secretariat is responsible for each of these areas and a Co-ordinator takes care of common problems. For the development of each of these documents a working group of a few world experts is first convened which prepare a preliminary draft. This draft is then reviewed by a larger, international Technical Review Committee (one for each of the five areas) and a subsequent review by the Senior Advisory Group - with representatives from 20 states - ensures that the document is well coordinated within the programme. At this stage, it is sent to Member States for comments. The Technical Review Committee concerned is reconvened to integrate these comments into the document, and, after a final review by the Senior Advisory Group, the document is ready for transmission to the Director General of the Agency for endorsement and publication. A preliminary to this procedure is the collation by the Secretariat of large amounts of information submitted by Member States so that the first draft is really based on a very complete knowledge of what is done in each area all over the world. This collation frequently reveals differences in approach which are not random but due, rather, to the local conditions and the types of reactors. These differences must be harmonized in the documents produced without detracting from the effectiveness of the code or guide. The whole

  2. Document reconstruction by layout analysis of snippets

    Science.gov (United States)

    Kleber, Florian; Diem, Markus; Sablatnig, Robert

    2010-02-01

    Document analysis is done to analyze entire forms (e.g. intelligent form analysis, table detection) or to describe the layout/structure of a document. Also skew detection of scanned documents is performed to support OCR algorithms that are sensitive to skew. In this paper document analysis is applied to snippets of torn documents to calculate features for the reconstruction. Documents can either be destroyed by the intention to make the printed content unavailable (e.g. tax fraud investigation, business crime) or due to time induced degeneration of ancient documents (e.g. bad storage conditions). Current reconstruction methods for manually torn documents deal with the shape, inpainting and texture synthesis techniques. In this paper the possibility of document analysis techniques of snippets to support the matching algorithm by considering additional features are shown. This implies a rotational analysis, a color analysis and a line detection. As a future work it is planned to extend the feature set with the paper type (blank, checked, lined), the type of the writing (handwritten vs. machine printed) and the text layout of a snippet (text size, line spacing). Preliminary results show that these pre-processing steps can be performed reliably on a real dataset consisting of 690 snippets.

  3. Studies relating to human intrusion into a repository. Report pertaining to work package 11. Preliminary safety case of the Gorleben site (VSG)

    Energy Technology Data Exchange (ETDEWEB)

    Beuth, Thomas; Buhmann, Dieter; Fischer-Appelt, Klaus; Moenig, Joerg; Ruebel, Andre; Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Bollingerfehr, Wilhelm; Filbert, Wolfgang [DBE Technology GmbH, Peine (Germany); Charlier, Frank [international nuclear safety engineering gmbh (nse), Aachen (Germany); Baltes, Bruno

    2014-10-15

    The question of the long-term safety of a repository system is inseparably linked with the intensive technical examination of the possible future evolution of the site and the repository system e. g. as a result of climatic, geologic, waste-related and repository-related processes. Here, the possible evolutions to be considered are those that have the potential to have a negative impact on the intended, furthest-possible, immediate, and lasting isolation of the radioactive waste in a defined area around the underground workings of the repository mine in salt rock, which is referred to as the containment-providing rock zone (CPRZ).

  4. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

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

  5. Web document engineering

    International Nuclear Information System (INIS)

    White, B.

    1996-05-01

    This tutorial provides an overview of several document engineering techniques which are applicable to the authoring of World Wide Web documents. It illustrates how pre-WWW hypertext research is applicable to the development of WWW information resources

  6. Enterprise Document Management

    Data.gov (United States)

    US Agency for International Development — The function of the operation is to provide e-Signature and document management support for Acquisition and Assisitance (A&A) documents including vouchers in...

  7. WIPP documentation plan

    International Nuclear Information System (INIS)

    Plung, D.L.; Montgomery, T.T.; Glasstetter, S.R.

    1986-01-01

    In support of the programs at the Waste Isolation Pilot Plant (WIPP), the Publications and Procedures Section developed a documentation plan that provides an integrated document hierarchy; further, this plan affords several unique features: 1) the format for procedures minimizes the writing responsibilities of the technical staff and maximizes use of the writing and editing staff; 2) review cycles have been structured to expedite the processing of documents; and 3) the numbers of documents needed to support the program have been appreciably reduced

  8. Documenting Employee Conduct

    Science.gov (United States)

    Dalton, Jason

    2009-01-01

    One of the best ways for a child care program to lose an employment-related lawsuit is failure to document the performance of its employees. Documentation of an employee's performance can provide evidence of an employment-related decision such as discipline, promotion, or discharge. When properly implemented, documentation of employee performance…

  9. ELECTRICAL POWER SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    M. Maniyar

    2004-06-22

    The purpose of this revision of the System Description Document (SDD) is to establish requirements that drive the design of the electrical power system and their bases to allow the design effort to proceed to License Application. This SDD is a living document that will be revised at strategic points as the design matures over time. This SDD identifies the requirements and describes the system design as they exist at this time, with emphasis on those attributes of the design provided to meet the requirements. This SDD has been developed to be an engineering tool for design control. Accordingly, the primary audience are design engineers. This type of SDD leads and follows the design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential to performing the design process. This SDD follows the design with regard to the description of the system. The description provided in the SDD is a reflection of the results of the design process to date. Functional and operational requirements applicable to this system are obtained from ''Project Functional and Operational Requirements'' (F&OR) (Siddoway, 2003). Other requirements to support the design process have been taken from higher level requirements documents such as ''Project Design Criteria Document'' (PDC) (Doraswamy 2004), the fire hazards analyses, and the preclosure safety analysis. The above mentioned low-level documents address ''Project Requirements Document'' (PRD) (Canori and Leitner 2003) requirements. This SDD includes several appendices with supporting information. Appendix B lists key system charts, diagrams, drawings, and lists; and Appendix C is a list of system procedures.

  10. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  11. Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    PIEPHO, M.G.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

  12. 29 CFR 1905.7 - Form of documents; subscription; copies.

    Science.gov (United States)

    2010-07-01

    ... UNDER THE WILLIAMS-STEIGER OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 General § 1905.7 Form of documents... 29 Labor 5 2010-07-01 2010-07-01 false Form of documents; subscription; copies. 1905.7 Section 1905.7 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION...

  13. Starlink Document Styles

    Science.gov (United States)

    Lawden, M. D.

    This document describes the various styles which are recommended for Starlink documents. It also explains how to use the templates which are provided by Starlink to help authors create documents in a standard style. This paper is concerned mainly with conveying the ``look and feel" of the various styles of Starlink document rather than describing the technical details of how to produce them. Other Starlink papers give recommendations for the detailed aspects of document production, design, layout, and typography. The only style that is likely to be used by most Starlink authors is the Standard style.

  14. Subject (of documents)

    DEFF Research Database (Denmark)

    Hjørland, Birger

    2017-01-01

    This article presents and discuss the concept “subject” or subject matter (of documents) as it has been examined in library and information science (LIS) for more than 100 years. Different theoretical positions are outlined and it is found that the most important distinction is between document......-oriented views versus request-oriented views. The document-oriented view conceive subject as something inherent in documents, whereas the request-oriented view (or the policy based view) understand subject as an attribution made to documents in order to facilitate certain uses of them. Related concepts...

  15. Effectiveness and safety of celecoxib combined with diacerein in the treatment of senile degenerative knee osteoarthritis: study protocol and preliminary results of a randomized controlled trial

    Directory of Open Access Journals (Sweden)

    Guo-dong Wang

    2017-01-01

    Full Text Available Background and objectives: The non-steroidal anti-inflammatory drug celecoxib, used for the treatment of senile degenerative knee osteoarthritis, has a rapid onset of action and few side effects. However, it cannot prevent the occurrence and development of knee osteoarthritis. Celecoxib withdrawal is likely to induce a rebound increase in inflammation that may aggravate symptoms. Diacerein is a drug used to improve articular cartilage metabolism. It can delay the progression of senile degenerative knee osteoarthritis, but has a slow onset of action. Whether combined treatment with celecoxib and diacerein provides complimentary actions to achieve a better therapeutic effect on senile degenerative knee osteoarthritis remains unclear. >Design: A prospective, single-center, randomized controlled trial. Methods: Three hundred patients with senile degenerative knee osteoarthritis who receive treatment at the Department of Orthopedics, Second Affiliated Hospital of Nanjing Medical University, China will be randomly assigned to the celecoxib (200 mg, once a day, orally, diacerein (50 mg, twice a day, orally, and celecoxib + diacerein (celecoxib 200 mg, once a day, orally and diacerein 50 mg, twice a day, orally groups, with 100 patients in each group. Patients in each group will be treated for 12 successive weeks, and a 36-week follow-up will be performed. At 24 weeks after treatment, the VAS score for pain during the 20-meter walk test and osteoarthritis index were significantly decreased in the diacerein and celecoxib + diacerein groups compared to pre-treatment (P Outcome measures and preliminary results: The primary outcome measure of this study is the Visual Analogue Scale (VAS score for pain during the 20-meter walk test before and 1, 4, 12, 24, and 36 weeks after treatment. The secondary outcome measures of this study include osteoarthritis index and X-ray image findings of the knee before and 1, 4, 12, 24, and 36 weeks after treatment, as well

  16. Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, P.O. BOX 1415, Zanjan (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, P.O. BOX 212 Tehran (Iran, Islamic Republic of); Kiai, S.M. Sadat [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I. P.O. BOX 14155-1339, Tehran (Iran, Islamic Republic of)], E-mail: sadatkiai@yahoo.com

    2009-02-15

    A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity ({rho}), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.

  17. Efficacy and safety of ultrasound-guided high intensity focused ultrasound ablation of symptomatic uterine fibroids in Black women: a preliminary study.

    Science.gov (United States)

    Zhang, C; Jacobson, H; Ngobese, Z E; Setzen, R

    2017-08-01

    To evaluate the therapeutic effect and safety of ultrasound-guided high-intensity focused ultrasound (USgHIFU) treatment on symptomatic uterine fibroids in Black women. A feasibility study. Gynaecological department in a teaching hospital in South Africa. Premenopausal women with uterus fibroids. Twenty-six patients with 53 fibroids who underwent USgHIFU treatment were enrolled. The USgHIFU treatment information was recorded, including treatment time, sonication time and total energy. Adverse events were also observed and recorded during and after treatment. Safety and efficacy of USgHIFU for the treatment of uterine fibroids in Black women. The median volume of fibroids was 52.7 (interquartile range, 18.6-177.4) cm 3 . According to USgHIFU treatment plan, total energy of 298.6 ± 169.3 kJ (range, 76.0-889.2) within treatment time of 90.3 ± 43.3 minutes (range, 14.0-208.0), in which sonication time of 774.0 ± 432.9 seconds (range, 190.0-2224.0) was used to ablate fibroids. The average ablation rate was 80.6 ± 9.7% (range, 46.5-94.5%). During the procedure, 69.2% of the patients reported lower abdominal pain, 57.7% sciatic/buttock pain, 38.5% burning skin, and 34.6% transient leg pain. No severe complications were observed. USgHIFU is feasible and safe to use to treat symptomatic uterine fibroids in Black women. Multiple uterine fibroids are more frequently detected in Black women. USgHIFU is feasible and safe for the treatment of uterine fibroids in Black women. © 2017 Royal College of Obstetricians and Gynaecologists.

  18. Safety and radiation-enhancing effect of sodium glycididazole in loco regionally advanced laryngeal cancers previously treated with platinum-containing chemotherapy regimens: a preliminary report

    International Nuclear Information System (INIS)

    Zeng, Y.C.; Wu, R.; Xu, Z.G.; Zhang, X.Y.; Wu, L.N.; Wang, Y.M.; Zheng, W.; Chen, X.D.; Chi, F.; Zhang, Z.Y.; Li, X.; Jin, X.Y.; Chen, W.; Wang, S.L.; Xiao, F.D.; Wang, E.Y.; Dong, X.Q.; Jia, M.X.; Li, Y.; Fan, G.L.; Hao, S.H.; Zhang, L.B.; Zhang, H.B.; Xia, H.H.X.

    2010-01-01

    Purpose: To determine the safety and radiation-enhancing effect of sodium glycididazole in laryngeal squamous cell carcinoma (stage T3-4,N0-3,M0) with conventional radiotherapy. Patients and methods: Patients with locoregional advanced laryngeal cancer (stage T3-4,N0-3,M0) were included: group 1(control, n = 30)were not administered of sodium glycididazole; group 2 (test, n = 30) received sodium glycididazole at a dose of 700 mg/m2 intravenous infusion 30 minutes before radiotherapy three times a week. Surrogate end-points of efficacy were tumor and nodal size. Safety parameters were vomiting, nausea, mucositis, laryngeal edema, esophagus and skin reaction, dysphagia, dyspnea, neurological deficit. Patients were evaluated weekly during treatment for 7 weeks and thereafter monthly for 3 months. Results: In the test, the overall response rate was 88.89% (95% CI, 71.00-97.00%) at 7 weeks and 92.59% (95% CI, 76.00 to 99.00%) at 1 month of follow-up. In the control, the overall response rate was 62.5% (95% CI, 41.00 to 81.00%) at 7 weeks and 58.33% (95% CI, 37.00 to 78.00%) at 1 month of follow-up. The short-term locoregional response rate was better in the test group at 7 weeks (p = 0.027) and at 1 month (p = 0.005) of follow-up. The test group had significantly more nausea and vomiting in weeks 1 (p = 0.047), 2 (p = 0.007), and 3 (p = 0.01) of treatment. Conclusions: The study indicates sodium glycididazole is an effective radiation-enhancing agent that improves short-term locoregional control and is well tolerated in patients with loco regionally advanced laryngeal cancer. (authors)

  19. SAFETY NOTES

    CERN Document Server

    TIS Secretariat

    2001-01-01

    Please note that the revisions of safety notes no 3 (NS 3 Rev. 2) and no 24 (NS 24 REV.) entitled respectively 'FIRE PREVENTION FOR ENCLOSED SPACES IN LARGE HALLS' and 'REMOVING UNBURIED ELV AND LVA ELECTRIC CONDUITS' are available on the web at the following urls: http://edmsoraweb.cern.ch:8001/cedar/doc.download?document_id=322811&version=1&filename=version_francaise.pdf http://edmsoraweb.cern.ch:8001/cedar/doc.download?document_id=322861&version=2&filename=version_francaise.pdf Paper copies can also be obtained from the TIS Divisional Secretariat, email tis.secretariat@cern.ch

  20. C-Band Airport Surface Communications System Engineering-Initial High-Level Safety Risk Assessment and Mitigation

    Science.gov (United States)

    Zelkin, Natalie; Henriksen, Stephen

    2011-01-01

    This document is being provided as part of ITT's NASA Glenn Research Center Aerospace Communication Systems Technical Support (ACSTS) contract: "New ATM Requirements--Future Communications, C-Band and L-Band Communications Standard Development." ITT has completed a safety hazard analysis providing a preliminary safety assessment for the proposed C-band (5091- to 5150-MHz) airport surface communication system. The assessment was performed following the guidelines outlined in the Federal Aviation Administration Safety Risk Management Guidance for System Acquisitions document. The safety analysis did not identify any hazards with an unacceptable risk, though a number of hazards with a medium risk were documented. This effort represents an initial high-level safety hazard analysis and notes the triggers for risk reassessment. A detailed safety hazards analysis is recommended as a follow-on activity to assess particular components of the C-band communication system after the profile is finalized and system rollout timing is determined. A security risk assessment has been performed by NASA as a parallel activity. While safety analysis is concerned with a prevention of accidental errors and failures, the security threat analysis focuses on deliberate attacks. Both processes identify the events that affect operation of the system; and from a safety perspective the security threats may present safety risks.

  1. Deep Borehole Disposal Safety Analysis.

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  2. TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS

    International Nuclear Information System (INIS)

    KRIPPS, L.J.

    2006-01-01

    This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for natural event hazard (NEH)-initiated accidents. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls

  3. Safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1976-01-01

    The Codes of Practice and Safety Guides that are being developed by the International Atomic Energy Agency are divided in five topical areas: Governmental Organization, Siting, Design, Operation and Quality Assurance. In each area, a scientific secretary is responsible for developing the documents and five Technical Review Committees composed of 10 to 12 experts from various Members Countries revise the drafts at different stages. A Senior Advisory Group supervises the entire programme and revises the document. A scientific co-ordinator is responsible for the co-ordination within the programme with other sections of the IAEA, and with other international organizations. In preparing a document, information on the practice adopted by Member States is collected, a group of experts is convened for preparing a preliminary draft on the basis of this material and the draft is then reviewed by the appropriate Technical Review Committee. The document is translated into various languages, reviewed by the Senior Advisory Group and sent to Member States for comments. After the comments of Member States have been received, the Technical Review Committee and then the Senior Advisory Group are convened again for the final revision of the document. Some 25 drafts, are in different stages of development. The preparation of a document in its final form takes about two years. The programme started in 1975 and to date most of the safety codes and a few safety guides have been sent to Member States for comments. These documents will have gone through the entire development procedure by early 1977. The Senior Advisory Groups and the Technical Review Committees meet on the average four times a year for a week at a time. Until now these meetings have been mainly concerned with the development of new documents or with that part of the procedure which precedes the transmission of the draft to Member States for comments. The next series of meetings will deal with the revisions needed to

  4. Preliminary Safety, Pharmacokinetics, and Efficacy of Regorafenib, Cisplatin, and Pemetrexed in Patients With Advanced Nonsquamous Non-Small-Cell Lung Cancers.

    Science.gov (United States)

    Hellmann, Matthew D; Sturm, Isrid; Trnkova, Zuzana Jirakova; Lettieri, John; Diefenbach, Konstanze; Rizvi, Naiyer A; Gettinger, Scott N

    2015-11-01

    Regorafenib is an oral multitargeted kinase inhibitor with potent antiangiogenic activity. In this phase I trial we evaluated the safety, pharmacokinetics, and efficacy of regorafenib with cisplatin and pemetrexed for patients with advanced nonsquamous non-small-cell lung cancers (nsNSCLCs). Nine patients enrolled before premature termination of the study. Five of 9 (56%) patients had a partial response and the median progression-free survival was 7 months (range, 1.5-15.1 months). Regorafenib had acceptable tolerability and minor pharmacokinetic interactions in combination with standard doses of cisplatin and pemetrexed in patients with advanced nsNSCLCs. The combination of bevacizumab, an antiangiogenesis agent, with cytotoxic chemotherapy improves survival in patients with advanced nonsquamous non-small-cell lung cancers (nsNSCLCs). Regorafenib is an oral multitargeted kinase inhibitor with potent antiangiogenic activity that is approved for patients with advanced colorectal cancer and gastrointestinal stromal tumors. In this phase I trial we evaluated the safety, pharmacokinetics (PK), and efficacy of regorafenib with cisplatin and pemetrexed for patients with advanced nsNSCLCs. Chemotherapy-naive patients with advanced nsNSCLCs were treated with regorafenib 60 mg/d continuously and cisplatin 75 mg/m(2) with pemetrexed 500 mg/m(2) once every 21 days for up to 6 cycles. Thereafter, regorafenib with or without pemetrexed could be continued as maintenance. Nine patients enrolled before premature termination of the study because of slow recruitment and a change in the development strategy of regorafenib by the study sponsor. Five patients experienced at least 1 treatment-related Grade 3 adverse event. No Grade 4 or 5 toxicity occurred. Five of 9 (56%) patients had a partial response and the median progression-free survival was 7 months (range, 1.5-15.1 months). Minor PK interactions between regorafenib and chemotherapy were observed. Regorafenib had acceptable

  5. Unreviewed safety question evaluation of 100 K West fuel canister gas and liquid sampling

    International Nuclear Information System (INIS)

    Alwardt, L.D.

    1995-01-01

    The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation for the gas and liquid sampling activities associated with the fuel characterization program at the 100 K West (KW) fuel storage basin. The scope of this safety evaluation is limited to the movement of canisters between the main storage basin, weasel pit, and south loadout pit transfer channel (also known as the decapping station); gas and liquid sampling of fuel canisters in the weasel pit; mobile laboratory preliminary sample analysis in or near the 105 KW basin building; and the placement of sample containers in an approved shipping container. It was concluded that the activities and potential accident consequences associated with the gas and liquid sampling of 100 KW fuel canisters are bounded by the current safety basis documents and do not constitute an Unreviewed Safety Question

  6. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  7. Scheme Program Documentation Tools

    DEFF Research Database (Denmark)

    Nørmark, Kurt

    2004-01-01

    are separate and intended for different documentation purposes they are related to each other in several ways. Both tools are based on XML languages for tool setup and for documentation authoring. In addition, both tools rely on the LAML framework which---in a systematic way---makes an XML language available...... as named functions in Scheme. Finally, the Scheme Elucidator is able to integrate SchemeDoc resources as part of an internal documentation resource....

  8. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    L.L. Swanson

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F andOR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts, diagrams, drawings, lists and additional supporting information; and Appendix C is a list of

  9. New Safety rules

    CERN Multimedia

    Safety Commission

    2008-01-01

    The revision of CERN Safety rules is in progress and the following new Safety rules have been issued on 15-04-2008: Safety Procedure SP-R1 Establishing, Updating and Publishing CERN Safety rules: http://cern.ch/safety-rules/SP-R1.htm; Safety Regulation SR-S Smoking at CERN: http://cern.ch/safety-rules/SR-S.htm; Safety Regulation SR-M Mechanical Equipment: http://cern.ch/safety-rules/SR-M.htm; General Safety Instruction GSI-M1 Standard Lifting Equipment: http://cern.ch/safety-rules/GSI-M1.htm; General Safety Instruction GSI-M2 Standard Pressure Equipment: http://cern.ch/safety-rules/GSI-M2.htm; General Safety Instruction GSI-M3 Special Mechanical Equipment: http://cern.ch/safety-rules/GSI-M3.htm. These documents apply to all persons under the Director General’s authority. All Safety rules are available at the web page: http://www.cern.ch/safety-rules The Safety Commission

  10. PRELIMINARY ENVIRONMENTAL, HEALTH AND SAFETY RISK ASSESSMENT ON THE INTEGRATION OF A PROCESS UTILIZING LOW-ENERGY SOLVENTS FOR CARBON DIOXIDE CAPTURE ENABLED BY A COMBINATION OF ENZYMES AND VACUUM REGENERATION WITH A SUBCRITICAL PC POWER PLANT

    Energy Technology Data Exchange (ETDEWEB)

    Fitzgerald, David; Vidal, Rafael; Russell, Tania; Babcock, Doosan; Freeman, Charles; Bearden, Mark; Whyatt, Greg; Liu, Kun; Frimpong, Reynolds; Lu, Kunlei; Salmon, Sonja; House, Alan; Yarborough, Erin

    2014-12-31

    The results of the preliminary environmental, health and safety (EH&S) risk assessment for an enzyme-activated potassium carbonate (K2CO3) solution post-combustion CO2 capture (PCC) plant, integrated with a subcritical pulverized coal (PC) power plant, are presented. The expected emissions during normal steady-state operation have been estimated utilizing models of the PCC plant developed in AspenTech’s AspenPlus® software, bench scale test results from the University of Kentucky, and industrial experience of emission results from a slipstream PCC plant utilizing amine based solvents. A review of all potential emission species and their sources was undertaken that identified two credible emission sources, the absorber off-gas that is vented to atmosphere via a stack and the waste removed from the PCC plant in the centrifuge used to reclaim enzyme and solvent. The conditions and compositions of the emissions were calculated and the potential EH&S effects were considered as well as legislative compliance requirements. Potential mitigation methods for emissions during normal operation have been proposed and solutions to mitigate uncontrolled releases of species have been considered. The potential emissions were found to pose no significant EH&S concerns and were compliant with the Federal legislation reviewed. The limitations in predicting full scale plant performance from bench scale tests have been noted and further work on a larger scale test unit is recommended to reduce the level of uncertainty.

  11. Health physics documentation

    International Nuclear Information System (INIS)

    Stablein, G.

    1980-01-01

    When dealing with radioactive material the health physicist receives innumerable papers and documents within the fields of researching, prosecuting, organizing and justifying radiation protection. Some of these papers are requested by the health physicist and some are required by law. The scope, quantity and deposit periods of the health physics documentation at the Karlsruhe Nuclear Research Center are presented and rationalizing methods discussed. The aim of this documentation should be the application of physics to accident prevention, i.e. documentation should protect those concerned and not the health physicist. (H.K.)

  12. CAED Document Repository

    Data.gov (United States)

    U.S. Environmental Protection Agency — Compliance Assurance and Enforcement Division Document Repository (CAEDDOCRESP) provides internal and external access of Inspection Records, Enforcement Actions, and...

  13. CFO Payment Document Management

    Data.gov (United States)

    US Agency for International Development — Paperless management will enable the CFO to create, store, and access various financial documents electronically. This capability will reduce time looking for...

  14. Efficacy and safety of Tofacitinib in patients with active rheumatoid arthritis resistant to conventional therapy: Preliminary results of an open-label clinical trial

    Directory of Open Access Journals (Sweden)

    E. L. Luchikhina

    2016-01-01

    Full Text Available Despite the advances in the therapy of rheumatoid arthritis (RA, which are associated with the use of biological anti-rheumatic drugs, the problemof effective treatment of RA is not still solved. Inclusion of new methods in treatment strategies, in particular the so-called «small molecules», i.e. synthetic compounds acting on intracellular signaling pathways, such as Tofacitinib (TOFA approved for use in rheumatologic practice, is very important.Objective: to evaluate the efficacy and safety of therapy with TOFA in combination with synthetic disease-modifying anti-rheumatic drugs (s-DMARDs, primarily methotrexate (MTX in in patients with active RA in real clinical practice.Subjects and methods. This ongoing open-label trial is a part of the scientific program «Russian Investigation of Methotrexate and Biologics in Early Active Inflammatory Arthritis» (REMARCA that explores the possibility of adapting the «treat-to-target» strategy in real prac-tice in Russia. The study included RA patients with moderate to high disease activity despite treatment with MTX or other DMARDs. A total of 41 patients with RA were included (8 males, 33 females; mean age 52.6±14.2 years, disease duration 47.2±49.7 months, 82.9% RF+ and 80.5% anti-CCP+,DAS28-ESR 5.45±0.95, SDAI 30.2±12.2. All the patients had previously received s-DMARDs; 12 (29.3% patients also had biological DMARDs (1 to 4 biologics. Oral TOFA 5 mg in combination with MTX or leflunomide was administered twice daily to 40 and 1 patients, respectively, with the possibility of increasing the dose up to 10 mg BID. To date, 37 and 12 patients received TOFA for 3 and 6 months, respectively.Results. TOFA was used as a second-line drug (after s-DMARDs failure in 29 (70.7%, as a third line drug (after s-DMARDs and biologics failure in 12 (29.3% patients. The dose was escalated to 10 mg BID in 13 (31.2% patients, on the average, 11.2±1.7 weeks after treatment initiation. TOFA was not

  15. Preliminary safety, pharmacokinetics, and efficacy of regorafenib, cisplatin, and pemetrexed in patients with advanced non-squamous non-small cell lung cancers

    Science.gov (United States)

    Hellmann, Matthew D; Sturm, Isrid; Trnkova, Zuzana Jirakova; Lettieri, John; Diefenbach, Konstanze; Rizvi, Naiyer A.; Gettinger, Scott N.

    2016-01-01

    Structured Abstract Purpose The addition of bevacizumab, an anti-angiogenesis agent, to cytotoxic chemotherapy improves survival in patients with advanced non-squamous non-small cell lung cancers (nsNSCLCs). Regorafenib is an oral multi-targeted kinase inhibitor with potent anti-angiogenic activity that is approved for patients with advanced colorectal cancer and gastrointestinal stromal tumors. This phase I trial evaluated the safety, pharmacokinetics, and efficacy of regorafenib with cisplatin and pemetrexed for patients with advanced nsNSCLCs. Methods Chemotherapy-naïve patients with advanced nsNSCLCs were treated with regorafenib 60mg/day continuously and cisplatin 75mg/m2 plus pemetrexed 500mg/m2 once every three weeks for up to six cycles. Thereafter, regorafenib with or without pemetrexed could be continued as maintenance. Results Nine patients enrolled prior to premature termination of the study due to slow recruitment and a change in the development strategy of regorafenib by the study sponsor, partially due to slow enrollment. Five patients experienced at least one treatment-related Grade 3 adverse event. No grade 4–5 toxicity occurred. 5 of 9 (56%) patients had a partial response and the median progression-free survival was 7 months (range 1.5–15.1). Minor pharmacokinetic (PK) interactions between regorafenib and chemotherapy were observed. Conclusions Regorafenib had acceptable tolerability and minor PK interactions in combination with standard doses of cisplatin and pemetrexed in patients with advanced nsNSCLCs. Encouraging activity was appreciated in chemotherapy-naïve patients with advanced nsNSCLCs. However, the small number of patients treated limits conclusions that can be drawn from these results. PMID:26003007

  16. Auditable safety analysis and final hazard classification for Buildings 1310-N and 1314-N

    International Nuclear Information System (INIS)

    Kloster, G.L.

    1997-05-01

    This document is a graded auditable safety analysis (ASA) of the deactivation activities planned for the 100-N facility segment comprised of the Building 1310-N pump silo (part of the Liquid Radioactive Waste Treatment Facility) and 1314-N Building (Liquid Waste Disposal Building).The ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the Final Hazard Classification (FHC) for the 1310-N pump silo and 1314-N Building segment. The FHC is radiological based on the Preliminary Hazard Classification and the total inventory of radioactive and hazardous materials in the segment

  17. Technetium removal: preliminary flowsheet options

    International Nuclear Information System (INIS)

    Eager, K.M.

    1995-01-01

    This document presents the results of a preliminary investigation into options for preliminary flowsheets for 99Tc removal from Hanford Site tank waste. A model is created to show the path of 99Tc through pretreatment to disposal. The Tank Waste Remediation (TWRS) flowsheet (Orme 1995) is used as a baseline. Ranges of important inputs to the model are developed, such as 99Tc inventory in the tanks and important splits through the TWRS flowsheet. Several technetium removal options are discussed along with sensitivities of the removal schemes to important model parameters

  18. Electronic document management at Sizewell B

    International Nuclear Information System (INIS)

    Rippon, Simon.

    1996-01-01

    Sizewell ''B'', Britain's first PWR, officially opened on March 25 1996, will now rely for its document management on a sophisticated computer-based system. One of the largest single engineering projects ever to be commissioned on one site in Britain, Sizewell ''B'' accommodates more than 300,000 documents, including over 200,000 drawings. The electronic document management system will provide a number of important benefits, including a more direct method of maintaining the station's Configuration Management and hence maintaining high safety standards; improved turnaround in plant modification proposals (PMP); significant time and cost savings in managing vital records; and increased productivity. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-08-31

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

  20. ITER-FEAT safety

    International Nuclear Information System (INIS)

    Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

    2001-01-01

    Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

  1. Safety first!

    CERN Multimedia

    2016-01-01

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

  2. Safety and preliminary efficacy data of a novel Casein Kinase 2 (CK2) peptide inhibitor administered intralesionally at four dose levels in patients with cervical malignancies

    International Nuclear Information System (INIS)

    Solares, Ana M; Alonso, Daniel F; Herrera, Luis; Sigman, Hugo; Perea, Silvio E; Acevedo, Boris E; López-Saura, Pedro; Santana, Agueda; Baladrón, Idania; Valenzuela, Carmen; González, Carlos A; Díaz, Alina; Castillo, Dagnelia; Ramos, Thelvia; Gómez, Roberto

    2009-01-01

    Cervical cancer is now considered the second leading cause of death among women worldwide, and its incidence has reached alarming levels, especially in developing countries. Similarly, high grade squamous intraepithelial lesion (HSIL), the precursor stage for cervical cancer, represents a growing health problem among younger women as the HSIL management regimes that have been developed are not fully effective. From the etiological point of view, the presence of Human Papillomavirus (HPV) has been demonstrated to play a crucial role for developing cervical malignancies, and viral DNA has been detected in 99.7% of cervical tumors at the later stages. CIGB-300 is a novel cyclic synthetic peptide that induces apoptosis in malignant cells and elicits antitumor activity in cancer animal models. CIGB-300 impairs the Casein Kinase (CK2) phosphorylation, by targeting the substrate's phosphoaceptor domain. Based on the perspectives of CIGB-300 to treat cancer, this 'first-in-human' study investigated its safety and tolerability in patients with cervical malignancies. Thirty-one women with colposcopically and histologically diagnosed microinvasive or pre-invasive cervical cancer were enrolled in a dose escalating study. CIGB-300 was administered sequentially at 14, 70, 245 and 490 mg by intralesional injections during 5 consecutive days to groups of 7 – 10 patients. Toxicity was monitored daily until fifteen days after the end of treatment, when patients underwent conization. Digital colposcopy, histology, and HPV status were also evaluated. No maximum-tolerated dose or dose-limiting toxicity was achieved. The most frequent local events were pain, bleeding, hematoma and erythema at the injection site. The systemic adverse events were rash, facial edema, itching, hot flashes, and localized cramps. 75% of the patients experienced a significant lesion reduction at colposcopy and 19% exhibited full histological regression. HPV DNA was negative in 48% of the

  3. IDC System Specification Document.

    Energy Technology Data Exchange (ETDEWEB)

    Clifford, David J.

    2014-12-01

    This document contains the system specifications derived to satisfy the system requirements found in the IDC System Requirements Document for the IDC Reengineering Phase 2 project. Revisions Version Date Author/Team Revision Description Authorized by V1.0 12/2014 IDC Reengineering Project Team Initial delivery M. Harris

  4. INFCE plenary conference documents

    International Nuclear Information System (INIS)

    This document consists of the reports to the First INFCE Plenary Conference (November 1978) by the Working Groups a Plenary Conference of its actions and decisions, the Communique of the Final INFCE Plenary Conference (February 1980), and a list of all documents in the IAEA depository for INFCE

  5. Human Document Project

    NARCIS (Netherlands)

    de Vries, Jeroen; Abelmann, Leon; Manz, A; Elwenspoek, Michael Curt

    2012-01-01

    “The Human Document Project‿ is a project which tries to answer all of the questions related to preserving information about the human race for tens of generations of humans to come or maybe even for a future intelligence which can emerge in the coming thousands of years. This document mainly

  6. Reactive documentation system

    Science.gov (United States)

    Boehnlein, Thomas R.; Kramb, Victoria

    2018-04-01

    Proper formal documentation of computer acquired NDE experimental data generated during research is critical to the longevity and usefulness of the data. Without documentation describing how and why the data was acquired, NDE research teams lose capability such as their ability to generate new information from previously collected data or provide adequate information so that their work can be replicated by others seeking to validate their research. Despite the critical nature of this issue, NDE data is still being generated in research labs without appropriate documentation. By generating documentation in series with data, equal priority is given to both activities during the research process. One way to achieve this is to use a reactive documentation system (RDS). RDS prompts an operator to document the data as it is generated rather than relying on the operator to decide when and what to document. This paper discusses how such a system can be implemented in a dynamic environment made up of in-house and third party NDE data acquisition systems without creating additional burden on the operator. The reactive documentation approach presented here is agnostic enough that the principles can be applied to any operator controlled, computer based, data acquisition system.

  7. Preliminary performance test of control rod position indicator for ballscrew type CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hu, H.; Lee, J. S.; Kim, J. I.

    2003-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The prototype of control rod position indicator having the high performance for the ballscrew type CEDM was developed on the basis of RSPT technology identified through the survey. The characteristics of control rod position indicator was defined and documented through design procedure and preliminary performance test

  8. Basis Document for Sludge Stabilization

    CERN Document Server

    Risenmay, H R

    2001-01-01

    DOE-RL recently issued Safety Evaluation Report (SER) amendments to the PFP Final Safety Analysis Report, HNF-SD-CP-SAR-021 Rev. 2. The Justification for Continued Operations for 2736-ZB and plutonium oxides in BTCs Safety Basis change (letter DOE-RL ABD-074) was approved by one of the SERs. Also approved by SER was the revised accident analysis for Magnesium Hydroxide Precipitation Process (MHPP) gloveboxes HC-230C-3 and HC-230C-5 containing increased glovebox inventories and corresponding increases in seismic release consequence. Numerous implementing documents require revision and issuance to implement the SER approvals. The SER plutonium oxides into BTCs specifically limited the SER scope to ''pure or clean oxides, i.e., 85 wt% or grater Pu, in this feed change'' (SER Section 3.0 Base Information paragraph 4 [page 11]). Comprehensive USQ Evaluation PFP-2001-12 addressed the packaging of Pu alloy metals into BTCs, and the packaging of Pu alloy oxides (powders) into food pack cans and determined that the ac...

  9. Preliminary assessments the shortcut to remediation (category III-surplus facility assessments)

    International Nuclear Information System (INIS)

    Byars, L.L.

    1995-01-01

    This report presents the details of the preliminary assessments for the shortcut of decontamination of surplus nuclear facilities. Topics discussed include: environment, health and safety concerns; economic considerations; reduction of transition time; preliminary characterization reports; preliminary project plan; health and safety plan; quality assurance plan; surveillance and maintenance plan; and waste management plan

  10. ISS Crew Transportation and Services Requirements Document

    Science.gov (United States)

    Bayt, Robert L. (Compiler); Lueders, Kathryn L. (Compiler)

    2016-01-01

    The ISS Crew Transportation and Services Requirements Document (CCT-REQ-1130) contains all technical, safety, and crew health medical requirements that are mandatory for achieving a Crew Transportation System Certification that will allow for International Space Station delivery and return of NASA crew and limited cargo. Previously approved on TN23183.

  11. Injury & Safety Report - Legacy

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Injury & Safety Report is a mandatory post trip legal document observers fill out to report any injuries they have incurred, illnesses they have had, or...

  12. Nuclear Safety in Central and Eastern Europe

    International Nuclear Information System (INIS)

    2001-04-01

    Nuclear safety is one of the critical issues with respect to the enlargement of the European Union towards the countries of Central and Eastern Europe. In the context of the enlargement process, the European Commission overall strategy on nuclear safety matters has been to bring the general standard of nuclear safety in the pre-accession countries up to a level that would be comparable to the safety levels in the countries of the European Union. In this context, the primary objective of the project was to develop a common format and general guidance for the evaluation of the current nuclear safety status in countries that operate commercial nuclear power plants. Therefore, one of the project team first undertakings was to develop an approach that would allow for a consistent and comprehensive overview of the nuclear safety status in the CEEC, enabling an equal treatment of the countries to be evaluated. Such an approach, which did not exist, should also ensure identification of the most important safety issues of the individual nuclear power plants. The efforts resulted in the development of the ''Performance Evaluation Guide'', which focuses on important nuclear safety issues such as plant design and operation, the practice of performing safety assessments, and nuclear legislation and regulation, in particular the role of the national regulatory body. Another important aspect of the project was the validation of the Performance Evaluation Guide (PEG) by performing a preliminary evaluation of nuclear safety in the CEEC, namely in Bulgaria, Czech Republic, Hungary, Lithuania, Romania, Slovak Republic, and Slovenia. The nuclear safety evaluation of each country was performed as a desktop exercise, using solely available documents that had been prepared by various Western institutions and the countries themselves. Therefore, the evaluation is only of a preliminary nature. The project did not intend to re-assess nuclear safety, but to focus on a comprehensive summary

  13. Status of safety analysis reports

    International Nuclear Information System (INIS)

    Cserhati, A.

    1999-01-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  14. Status of safety analysis reports

    Energy Technology Data Exchange (ETDEWEB)

    Cserhati, A

    1999-06-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  15. Strategic environmental safety inspection for the National disposal program. Description of the inspection volume. Documentation for the scoping team; Strategische Umweltpruefung zum Nationalen Entsorgungsprogramm. Beschreibung des Untersuchungsumfangs. Unterlage fuer den Scoping-Termin

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-01-06

    The Strategic environmental safety inspection for the National disposal program covers the following topics: Legal framework: determination of the requirement for an environmental inspection program, coordination of the scoping team into the overall context; environmental targets; approach for assessment and evaluation of environmental impact, description of the inspection targets for the strategic environmental inspection; consideration of alternatives.

  16. TRANSPORTATION SYSTEM REQUIREMENTS DOCUMENT

    International Nuclear Information System (INIS)

    2004-01-01

    This document establishes the Transportation system requirements for the U.S. Department of Energy's (DOE's) Civilian Radioactive Waste Management System (CRWMS). These requirements are derived from the Civilian Radioactive Waste Management System Requirements Document (CRD). The Transportation System Requirements Document (TSRD) was developed in accordance with LP-3.1Q-OCRWM, Preparation, Review, and Approval of Office of National Transportation Level-2 Baseline Requirements. As illustrated in Figure 1, the TSRD forms a part of the DOE Office of Civilian Radioactive Waste Management (OCRWM) Technical Baseline

  17. Fiscal year 1999 waste information requirements document

    International Nuclear Information System (INIS)

    Adams, M.R.

    1998-01-01

    The Waste Information Requirements Document (WIRD) has the following purposes: To describe the overall drivers that require characterization information and to document their source; To define how characterization is going to satisfy the drivers, close issues, and measure and report progress; and To describe deliverables and acceptance criteria for characterization. Characterization information is required to maintain regulatory compliance, perform operations and maintenance, resolve safety issues, and prepare for disposal of waste. Commitments addressing these requirements are derived from the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement; the Recommendation 93-5 Implementation Plan (DOE-RL 1996a) to the Defense Nuclear Facilities Safety Board (DNFSB); and other requirement sources listed in Section 2.0. The Waste Information Requirements Document replaces the tank waste analysis plans and the tank characterization plan previously required by the Tri-Party Agreement, Milestone M-44-01 and M-44-02 series

  18. Safety aspects in life extension of NPPs. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    Due to current social and economical framework, in last years many Member States (MS) started a process of Plant Life Extension (PLEX) for their older nuclear facilities. The process followed many different approaches, being intrinsically dependent on the national regulatory framework and technical tradition. This process has many nuclear safety implications, other than strategic and political ones, and therefore a need for tailoring the available safety assessment tools to such applications has become urgent in recent years. Typical safety assessment processes such as the Periodic Safety Review has been used already and can be used in the future as a framework for a PLEX. Also the review of regular maintenance and ageing management programs and the continuous upgrading of the Safety Analysis Report are tools widely used in a life extension context in many MS. However, recently some MS highlighted the need to identify in a clear way the technical aspects more directly affecting the decision for a long-term operation of a nuclear facility. Many Technical Cooperation projects have been requested for the year 2003-2004 on this subject and a generic task in Nuclear Safety proved necessary as a background activity. Therefore, on May 6-10, 2002, a Consultant Meeting dealing with Safety Aspects of life extension for NPPs was convened at the Nuclear Safety Dept. of the IAEA. It was attended by representatives of Regulatory Bodies and Utilities, both from countries with experience of life extension of NPPs and countries where the process is at the beginning. The main application problems were identified and discussed and a first attempt was carried out to define the key elements of the life extension process, isolating peculiar technical items, LTO related, from generic safety related tasks. The result was a preliminary technical document with a collection of basic experience and information for the implementation of a PLEX program. Therefore the draft document was thought

  19. Exploratory shaft facility preliminary designs - Gulf Interior Region salt domes

    International Nuclear Information System (INIS)

    1983-09-01

    The purpose of the Preliminary Design Report, Gulf Interior Region, is to provide a description of the preliminary design for an Exploratory Shaft Facility on the Richton Dome, Mississippi. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers are included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description and Construction Cost Estimate

  20. Exploratory shaft facility preliminary designs - Paradox Basin. Technical report

    International Nuclear Information System (INIS)

    1983-09-01

    The purpose of the Preliminary Design Report, Paradox Basin, is to provide a description of the preliminary design for an Exploratory Shaft Facility in the Paradox Basin, Utah. This issue of the report describes the preliminary design for constructing the exploratory shaft using the Large Hole Drilling Method of construction and outlines the preliminary design and estimates of probable construction cost. The Preliminary Design Report is prepared to complement and summarize other documents that comprise the design at the preliminary stage of completion, December 1982. Other design documents include drawings, cost estimates and schedules. The preliminary design drawing package, which includes the construction schedule drawing, depicts the descriptions in this report. For reference, a list of the drawing titles and corresponding numbers is included in the Appendix. The report is divided into three principal sections: Design Basis, Facility Description, and Construction Cost Estimate. 30 references

  1. Informational system. Documents management

    Directory of Open Access Journals (Sweden)

    Vladut Iacob

    2009-12-01

    Full Text Available Productivity growing, as well as reducing of operational costs in a company can be achieved by adopting a document management solutions. Such application will allow management and structured and efficient transmission of information within the organization.

  2. Transportation System Requirements Document

    International Nuclear Information System (INIS)

    1993-09-01

    This Transportation System Requirements Document (Trans-SRD) describes the functions to be performed by and the technical requirements for the Transportation System to transport spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from Purchaser and Producer sites to a Civilian Radioactive Waste Management System (CRWMS) site, and between CRWMS sites. The purpose of this document is to define the system-level requirements for Transportation consistent with the CRWMS Requirement Document (CRD). These requirements include design and operations requirements to the extent they impact on the development of the physical segments of Transportation. The document also presents an overall description of Transportation, its functions, its segments, and the requirements allocated to the segments and the system-level interfaces with Transportation. The interface identification and description are published in the CRWMS Interface Specification

  3. Integrated Criteria Document Chromium

    NARCIS (Netherlands)

    Slooff W; Cleven RFMJ; Janus JA; van der Poel P; van Beelen P; Boumans LJM; Canton JH; Eerens HC; Krajnc EI; de Leeuw FAAM; Matthijsen AJCM; van de Meent D; van der Meulen A; Mohn GR; Wijland GC; de Bruijn PJ; van Keulen A; Verburgh JJ; van der Woerd KF

    1990-01-01

    Betreft de engelse versie van rapport 758701001
    Bij dit rapport behoort een appendix onder hetzelfde nummer getiteld: "Integrated Criteria Document Chromium: Effects" Auteurs: Janus JA; Krajnc EI
    (appendix: see 710401002A)

  4. NCDC Archive Documentation Manuals

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The National Climatic Data Center Tape Deck Documentation library is a collection of over 400 manuals describing NCDC's digital holdings (both historic and current)....

  5. Registration document 2005

    International Nuclear Information System (INIS)

    2005-01-01

    This reference document of Gaz de France provides information and data on the Group activities in 2005: financial informations, business, activities, equipments factories and real estate, trade, capital, organization charts, employment, contracts and research programs. (A.L.B.)

  6. Are PDF Documents Accessible?

    Directory of Open Access Journals (Sweden)

    Mireia Ribera Turró

    2008-09-01

    Full Text Available Adobe PDF is one of the most widely used formats in scientific communications and in administrative documents. In its latest versions it has incorporated structural tags and improvements that increase its level of accessibility. This article reviews the concept of accessibility in the reading of digital documents and evaluates the accessibility of PDF according to the most widely established standards.

  7. 2002 reference document

    International Nuclear Information System (INIS)

    2002-01-01

    This 2002 reference document of the group Areva, provides information on the society. Organized in seven chapters, it presents the persons responsible for the reference document and for auditing the financial statements, information pertaining to the transaction, general information on the company and share capital, information on company operation, changes and future prospects, assets, financial position, financial performance, information on company management and executive board and supervisory board, recent developments and future prospects. (A.L.B.)

  8. LCS Content Document Application

    Science.gov (United States)

    Hochstadt, Jake

    2011-01-01

    My project at KSC during my spring 2011 internship was to develop a Ruby on Rails application to manage Content Documents..A Content Document is a collection of documents and information that describes what software is installed on a Launch Control System Computer. It's important for us to make sure the tools we use everyday are secure, up-to-date, and properly licensed. Previously, keeping track of the information was done by Excel and Word files between different personnel. The goal of the new application is to be able to manage and access the Content Documents through a single database backed web application. Our LCS team will benefit greatly with this app. Admin's will be able to login securely to keep track and update the software installed on each computer in a timely manner. We also included exportability such as attaching additional documents that can be downloaded from the web application. The finished application will ease the process of managing Content Documents while streamlining the procedure. Ruby on Rails is a very powerful programming language and I am grateful to have the opportunity to build this application.

  9. Documentation of spectrom-32

    International Nuclear Information System (INIS)

    Callahan, G.D.; Fossum, A.F.; Svalstad, D.K.

    1989-01-01

    SPECTROM-32 is a finite element program for analyzing two-dimensional and axisymmetric inelastic thermomechanical problems related to the geological disposal of nuclear waste. The code is part of the SPECTROM series of special-purpose computer programs that are being developed by RE/SPEC Inc. to address many unique rock mechanics problems encountered in analyzing radioactive wastes stored in geologic formations. This document presents the theoretical basis for the mathematical models, the finite element formulation and solution procedure of the program, a description of the input data for the program, verification problems, and details about program support and continuing documentation. The computer code documentation is intended to satisfy the requirements and guidelines outlined in the document entitled Final Technical Position on Documentation of Computer Codes for High-Level Waste Management. The principal component models used in the program involve thermoelastic, thermoviscoelastic, thermoelastic-plastic, and thermoviscoplastic types of material behavior. Special material considerations provide for the incorporation of limited-tension material behavior and consideration of jointed material behavior. Numerous program options provide the capabilities for various boundary conditions, sliding interfaces, excavation, backfill, arbitrary initial stresses, multiple material domains, load incrementation, plotting database storage and access of results, and other features unique to the geologic disposal of radioactive wastes. Numerous verification problems that exercise many of the program options and illustrate the required data input and printed results are included in the documentation

  10. Technical approach document

    International Nuclear Information System (INIS)

    1988-04-01

    This document describes the general technical approaches and design criteria adopted by the US Department of Energy (DOE) in order to implement Remedial Action Plans (RAPs) and final designs that comply with EPS standards. This document is a revision to the original document. Major revisions were made to the sections in riprap selection and sizing, and ground-water; only minor revisions were made to the remainder of the document. The US Nuclear Regulatory Commission (NRC) has prepared a Standard Review Plan (NRC-SRP) which describes factors to be considered by the NRC in approving the RAP. Sections 3.0, 4.0, 5.0, and 7.0 of this document are arranged under the same headings as those used in the NRC-SRP. This approach is adopted in order to facilitate joint use of the documents. Section 2.0 (not included in the NRC-SRP) discusses design considerations; Section 3.0 describes surface-water hydrology and erosion control; Section 4.0 describes geotechnical aspects of pile design; Section 5.0 discusses the Alternate Site Selection Process; Section 6.0 deals with radiological issues (in particular, the design of the radon barrier); Section 7.0 discusses protection of groundwater resources; and Section 8.0 discusses site design criteria for the RAC

  11. Safety Basis Report

    International Nuclear Information System (INIS)

    R.J. Garrett

    2002-01-01

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities

  12. Safety Basis Report

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2002-01-14

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.

  13. AVLIS Production Plant Preliminary Quality Assurance Plan and Assessment

    International Nuclear Information System (INIS)

    1984-01-01

    This preliminary Quality Assurance Plan and Assessment establishes the Quality Assurance requirements for the AVLIS Production Plant Project. The Quality Assurance Plan defines the management approach, organization, interfaces, and controls that will be used in order to provide adequate confidence that the AVLIS Production Plant design, procurement, construction, fabrication, installation, start-up, and operation are accomplished within established goals and objectives. The Quality Assurance Program defined in this document includes a system for assessing those elements of the project whose failure would have a significant impact on safety, environment, schedule, cost, or overall plant objectives. As elements of the project are assessed, classifications are provided to establish and assure that special actions are defined which will eliminate or reduce the probability of occurrence or control the consequences of failure. 8 figures, 18 tables

  14. Preliminary hazard classification for buildings 1310-N and 1314-N

    International Nuclear Information System (INIS)

    Kloster, G.L.; Smith, R.I.

    1997-01-01

    This document establishes the preliminary hazard classification (PHC) for the 100-N Area facility segment comprised of the 1310-N ''silo'' building and the 1314-N Liquid Waste Disposal Building. To establish the PHC, the inventories of radioactive and nonradioactive hazardous materials present within the segment are identified and then compared to the corresponding threshold quantity values in DOE-STD-1027-92 and reportable quantity values in 40 CFR 302.4. In this evaluation, no credit is taken for the form, location, and dispersibility of the materials; for their interaction with available energy sources; or for safety features that could prevent or mitigate a radiological release. The result of the PHC determined that the 1310-N and 1314-N building segments are classified as radiological

  15. The application of integrated safety management principles to the Tritium Extraction Facility project

    International Nuclear Information System (INIS)

    Hickman, M.O.; Viviano, R.R.

    2000-01-01

    The DOE has developed a program that is accomplishing a heightened safety posture across the complex. The Integrated Safety Management (ISM) System (ISMS) program utilizes five core functions and seven guiding principles as the basis for implementation. The core functions define the work scope, analyze the hazards, develop and implement hazard controls, perform the work, and provide feedback for improvement. The guiding principles include line management responsibility, clear roles and responsibilities, competence per responsibilities, identification of safety standards/requirements, tailored hazard control, balanced priorities, and operations authorization. There exists an unspecified eighth principle, that is, worker involvement. A program requiring the direct involvement of the employees who are actually performing the work has been shown to be quite an effective method of communicating safety requirements, controlling work in a safe manner, and reducing safety violations and injuries. The Tritium Extraction Facility (TEF) projects, a component of the DOE's Commercial Light Water Reactor Tritium Production program, has taken the ISM principles and core functions and applied them to the project's design. The task of the design team is to design a facility and systems that will meet the production requirements of the DOE tritium mission as well as a design that minimizes the workers' exposure to adverse safety situations and hazards/hazardous materials. During the development of the preliminary design for the TEF, design teams consisted of not only designers but also personnel who had operational experience in the existing tritium and personnel who had operational experience in the existing tritium and personnel who had specialized experience from across the DOE complex. This design team reviewed multiple documents associated with the TEF operation in order to identify and document the hazards associated with the tritium process. These documents include hazards

  16. Criticality safety evaluation report for FFTF 42% fuel assemblies

    International Nuclear Information System (INIS)

    Richard, R.F.

    1997-01-01

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC)

  17. Guidelines for the review of accident management programmes in nuclear power plants. Reference document for the IAEA safety service missions on review of accident management programmes in nuclear power plants

    International Nuclear Information System (INIS)

    2003-01-01

    Similarly as for other IAEA safety services, the objectives of accident management safety service are to assist the Member States in ensuring and enhancing the safety of NPPs. In particular, the objective is to assist at the utility and NPP (i.e. licensee) level in effective plant specific AMP preparation, development and implementation. However, assistance can also be provided to the regulatory body in its reviewing of AMPs. Objectives of the safety service can be summarized as follows: To explain to licensee personnel principles and possible approaches in effective implementation of AMP based on experience world-wide; To give opportunities to experts from the host plant to broaden their experience and knowledge in the field; To perform an objective assessment of the status in various phases of AMP implementation, compared with international experience and practices; To provide the licensee with suggestions and assistance for improvements in various stages of AMP implementation. The objective of the IAEA safety services is to offer two options to respond to individual requirements. These options include missions to review accident analysis needed for accident management and missions to review the whole AMP. Review of accident analysis for accident management (RAAAM): this review is intended to check completeness and quality of accident analysis covering BDBA and severe accidents. The review should be typically performed prior to use of accident analysis for development of AMP. It is considered that 2 experts and 1 IAEA team leader in one-week mission can perform the review. Detailed guidelines for review of analysis are provided in Section 2. Reference is also made to another IAEA Safety Report (Safety Standards Series No. NS-R-1) which is devoted to guidance for accident analysis of nuclear power plants (NPPs). Review of AMP (RAMP): this review of AMP, which is in particular appropriate prior to its implementation, is intended to check its quality, consistency

  18. The European Utility Requirement Document

    International Nuclear Information System (INIS)

    Roche, I.I.

    1999-01-01

    The major European electricity producers work on a common requirement document for future LWR plants since 1992. They aim at requirements acceptable together by the owners, the public and the authorities. Thus the designers can develop standard LWR designs acceptable everywhere in Europe and the utilities can open their consultations to vendors on common bases. Such a standardisation promotes an improvement of generation costs and of safety : public and authorities acceptance should be improved as well ; significant savings are expected in development and construction costs. Since the early stages of the project, the EUR group has grown significantly. It now includes utilities from nine European countries. Utilities from two other European countries are joining the group. Specific cooperation agreements are also in progress with a few extra-European partners

  19. Safety of installations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-09-01

    This document presents the fulfilling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 5 of the document contains some details about the siting, designed and construction, operation, maintenance, inspection, testing, procedures for responding to anticipated operational occurrences and accidents, engineering and technical support, reporting of significant incidents, operating experience feedback, radioactive waste and spent fuel.

  20. Safety of installations

    International Nuclear Information System (INIS)

    2001-01-01

    This document presents the fulfilling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 5 of the document contains some details about the siting, designed and construction, operation, maintenance, inspection, testing, procedures for responding to anticipated operational occurrences and accidents, engineering and technical support, reporting of significant incidents, operating experience feedback, radioactive waste and spent fuel