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)

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

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

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)

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

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

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

[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.

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

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.

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

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)

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

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

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

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.

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

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

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

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)

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.

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

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

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.

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

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)

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.

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

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.

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.

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.

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)

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

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

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

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

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

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.

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

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

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.

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

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

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...

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

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.)

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)

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

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.

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.

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

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.

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

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'

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

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

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

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

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

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.

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

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.)

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.

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

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)

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

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.

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

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)

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

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)

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

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).

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

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.

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.

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

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

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.

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

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

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

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

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

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

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

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

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

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)

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

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.

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.)

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

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

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.

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.

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

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.

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).

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.

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)

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

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.

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.

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.

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

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.

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

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)

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...

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.

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.

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

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

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

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.

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.

Requirement analysis of the safety-critical software implementation for the nuclear power plant

International Nuclear Information System (INIS)

Chang, Hoon Seon; Jung, Jae Cheon; Kim, Jae Hack; Nam, Sang Ku; Kim, Hang Bae

2005-01-01

The safety critical software shall be implemented under the strict regulation and standards along with hardware qualification. In general, the safety critical software has been implemented using functional block language (FBL) and structured language like C in the real project. Software design shall comply with such characteristics as; modularity, simplicity, minimizing the use of sub-routine, and excluding the interrupt logic. To meet these prerequisites, we used the computer-aided software engineering (CASE) tool to substantiate the requirements traceability matrix that were manually developed using Word processors or Spreadsheets. And the coding standard and manual have been developed to confirm the quality of software development process, such as; readability, consistency, and maintainability in compliance with NUREG/CR-6463. System level preliminary hazard analysis (PHA) is performed by analyzing preliminary safety analysis report (PSAR) and FMEA document. The modularity concept is effectively implemented for the overall module configurations and functions using RTP software development tool. The response time imposed on the basis of the deterministic structure of the safety-critical software was measured

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

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)

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...

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.

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)

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.

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

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.

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

Safety culture and learning from incidents: the role of incident reporting and causal analyses

International Nuclear Information System (INIS)

Wilpert, B.

1994-01-01

Nuclear industry more than any other industrial branch has developed and used predictive risk analysis as a method of feedforward control of safety and reliability. Systematic evaluation of operating experience, statistical documentation of component failures, systematic documentation and analysis of incidents are important complementary elements of feedback control: we are dealing here with adjustment and learning from experience, in particular from past incidents. Using preliminary findings from ongoing research at the Research Center Systems Safety at the Berlin University of Technology the contribution discusses preconditions for an effective use of lessons to be learnt from closely matched incident reporting and in depth analyses of causal chains leading to incidents. Such conditions are especially standardized documentation, reporting and analyzing methods of incidents; structured information flows and feedback loops; abstaining from culpability search; mutual trust of employees and management; willingness of all concerned to continually evaluate and optimize the established learning system. Thus, incident related reporting and causal analyses contribute to safety culture, which is seen to emerge from tightly coupled organizational measures and respective change in attitudes and behaviour. (author) 2 figs., 7 refs

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

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)

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

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

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.)

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

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.

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

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.

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.

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

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

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

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)

Hazard classification and auditable safety analysis for the 1300-N Emergency Dump Basin

International Nuclear Information System (INIS)

Kretzschmar, S.P.; Larson, A.R.

1996-06-01

This document combines the following four analytical functions: (1) hazards baseline of the Emergency Dump Basin (EDB) in the quiescent state; (2) preliminary hazard classification for intrusive activities (i.e., basin stabilization); (3) final hazard classification for intrusive activities; and (4) an auditable safety analysis. This document describes the potential hazards contained within the EDB at the N Reactor complex and the vulnerabilities of those hazards during the quiescent state (when only surveillance and maintenance activities take place) and during basin stabilization activities. This document also identifies the inventory of both radioactive and hazardous material in the EDB. Result is that the final hazard classification for the EDB segment intrusive activities is radiological

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

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

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

Obtaining Valid Safety Data for Software Safety Measurement and Process Improvement

Science.gov (United States)

Basili, Victor r.; Zelkowitz, Marvin V.; Layman, Lucas; Dangle, Kathleen; Diep, Madeline

2010-01-01

We report on a preliminary case study to examine software safety risk in the early design phase of the NASA Constellation spaceflight program. Our goal is to provide NASA quality assurance managers with information regarding the ongoing state of software safety across the program. We examined 154 hazard reports created during the preliminary design phase of three major flight hardware systems within the Constellation program. Our purpose was two-fold: 1) to quantify the relative importance of software with respect to system safety; and 2) to identify potential risks due to incorrect application of the safety process, deficiencies in the safety process, or the lack of a defined process. One early outcome of this work was to show that there are structural deficiencies in collecting valid safety data that make software safety different from hardware safety. In our conclusions we present some of these deficiencies.

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

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

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)

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

Preliminary Hazard Classification for the 105-B Reactor

International Nuclear Information System (INIS)

Kerr, N.R.

1997-08-01

This document summarizes the inventories of radioactive and hazardous materials present within the 105-B Reactor and uses the inventory information to determine the preliminary hazard classification for the surveillance and maintenance activities of the facility. The result of this effort was the preliminary hazard classification for the 105-B Building surveillance and maintenance activities. The preliminary hazard classification was determined to be Nuclear Category 3. Additional hazard and accident analysis will be documented in a separate report to define the hazard controls and final hazard classification

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

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.

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

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

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

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.

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

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.

Development of safety analysis technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

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?

Preliminary design report: Prototypical Spent Fuel Consolidation Equipment Demonstration Project: Phase 1

International Nuclear Information System (INIS)

Blissell, W.H.; Ciez, A.P.; Mitchell, J.L.; Winkler, C.J.

1986-12-01

This document describes the Westinghouse Preliminary Design for the Prototypical Consolidation Demonstration Project per Department of Energy (DOE) Contract No. DE-AC07-86ID12649 and under direction of the DOE Idaho Operations Office. The preliminary design is the first step to providing the Department of Energy with a fully qualified, licensable, cost-effective spent fuel rod consolidation system. The design was developed using proven technologies and equipment to create an innovative approach to previous rod consolidation concepts. These innovations will better enable the Westinghouse system to: consolidate fuel rods in a precise, fully-controlled, accountable manner; package all rods from two PWR fuel assemblies or from four BWR fuel assemblies in one 8.5 inch square consolidated rods canister; meet all functional requirements; operate with all fuel types common to the US commercial nuclear industry with minimal tooling changeouts; and meet consolidation production process rates, while maintaining operator and public health and safety. This Preliminary Design Report provides both detailed descriptions of the equipment required to perform the rod consolidation process and the supporting analyses to validate the design

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

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.

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

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Jang-Soo; Jee, Eunkyoung

2016-01-01

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee-Choon; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jee, Eunkyoung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents.

Preparation of the initial safety case

International Nuclear Information System (INIS)

Hensley, G.

1987-01-01

In British Nuclear Fuels plc (BNFL), the design of nuclear chemical plants for construction and subsequent operation at Sellafield Works is carried out by the Engineering Division of the Spent Fuel Management Services Group based at Risley, Warrington. Plant construction cannot take place, nor plant commissioning, until it has been demonstrated in the initial (design) safety case that the chosen design will allow the plant to be operated in an adequately safe manner, corresponding to an extremely low level of risk. The safety documentation procedure is described. A Preliminary Design Safety Appraisal is made of the initial design proposal to give an early indication of the order of risk that might prevail. The risk from each hazard is compared with an allocated risk target which makes up a proportion of the total plant risk which is quantified in BNFL's risk criteria. Where the risk appears unacceptable, appropriate modifications are made to the design. Prior to commissioning, a comprehensive, detailed risk assessment is carried out. The methodology of probabilistic risk assessment is described and examples given of how different hazards are assessed. (author)

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.

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

Development of guidance and methodical documents for providing the decommissioning of radiation-hazardous objects

International Nuclear Information System (INIS)

Ermakov, A.

2015-01-01

Federal Center for Nuclear and Radiation Safety (JSC FCNRS) developed and approbated guidance and methodical documents for providing the facility to radiologically safe status in the course of Building B decommissioning activity at JSC VNIINM (A.A. Bochvar High-Technology Scientific Research Institute for Inorganic Materials). The scope of application of the documents developed is as follows: - preliminary segregation of radwaste into streams during its collection and preparation for removal from facilities/sites under decommissioning; - express assessment of specific activity (activity) of radwaste generated in the course of dismantling and decontamination activities; - radiological survey of premises and building structures following completion of dismantling and decontamination activities; - SRW processing (compaction, reduction in size), packaging, characterisation and containerisation in order to reduce risks of spread of radioactive contamination. Documents that have been developed can be used both at nuclear facilities/ sites similar to the JSC VNIINM Building B in terms of work stages and types of waste to be generated, and other facilities/ sites taking into consideration their peculiarities. (author)

IAEA safety fundamentals: the safety of nuclear installations and the defence in depth concept

International Nuclear Information System (INIS)

Aro, I.

2005-01-01

This presentation is a replica of the similar presentation provided by the IAEA Basic Professional Training Course on Nuclear Safety. The presentation utilizes the IAEA Safety Series document No. 110, Safety Fundamentals: the Safety of Nuclear Installations. The objective of the presentation is to provide the basic rationale for actions in provision of nuclear safety. The presentation also provides basis to understand national nuclear safety requirements. There are three Safety Fundamentals documents in the IAEA Safety Series: one for nuclear safety, one for radiation safety and one for waste safety. The IAEA is currently revising its Safety Fundamentals by combining them into one general Safety Fundamentals document. The IAEA Safety Fundamentals are not binding requirements to the Member States. But, a very similar text has been provided in the Convention on Nuclear Safety which is legally binding for the Member State after ratification by the Parliament. This presentation concentrates on nuclear safety. The Safety Fundamentals documents are the 'policy documents' of the IAEA Safety Standards Series. They state the basic objectives, concepts and principles involved in ensuring protection and safety in the development and application of atomic energy for peaceful purposes. They will state - without providing technical details and without going into the application of principles - the rationale for actions necessary in meeting Safety Requirements. Chapter 7 of this presentation describes the basic features of defence in depth concept which is referred to in the Safety Fundamentals document. The defence in depth concept is a key issue in reaching high level of safety specifically at the design stage but as the reader can see the extended concept also refers to the operational stage. The appendix has been taken directly from the IAEA Basic Professional Training Course on Nuclear Safety and applied to the Finnish conditions. The text originates from the references

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.

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

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

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

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)

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.

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

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.

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

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...

New standard on safety for hydrogen systems in spanish. Keys for understanding and use

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-15

The present paper approaches all the preliminary, normative and additional elements observed during the work carried out by the Argentine standardization board to count in the country with a normative document that covers the expectations of the local community of users and other Spanish-speaking user, about the integral safety for the hydrogen systems. The antecedents and the process of adoption of an international standard and its adaptation to the local media are analyzed. The result has been the Standard IRAM/ISO 15916 that intends to offer, to all the users and especially to those who are not familiar with the technology, a base to understand the subject of safety, thus enhancing the education of the general public in hydrogen safety matters. (author)

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

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

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.

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

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

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

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

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

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

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.

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

Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

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

Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for 330MWt SMART integral reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, B. D.; Lee, W. J.; Sim, S. K.; Song, J. H.; Kim, H. C.

1997-09-01

The work reported in this document identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in a 330 MWt SMART integral reactor which is under development at KAERI. The result of this efforts is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by the consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this report is intended for use to identify and integrate development areas of further experimental tests needed and thermal-hydraulic models and correlations and code improvements for the safety analysis of the SMART integral reactor. (author). 7 refs., 21 tabs., 22 figs.

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.

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

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.

Management of Documents and Information in BSC Secretariat

International Nuclear Information System (INIS)

Sumathi, E.; Jayarajan, K.

2017-01-01

The regulatory and safety function of BARC facilities is being carried out by BARC Safety Framework with BARC Safety Council as an apex body. Presently, about one hundred safety committees and task forces are functional in the framework. BSC Secretariat (BSCS) provides technical and administrative support to the BARC Safety Framework for the regulatory activities in BARC. Important documents/records related to committee decisions and facilities are maintained in BSCS, through an established documentation and record keeping system. The compliance of regulatory recommendations is verified during regulatory inspections and subsequent submissions made by the facility authority. This supports the effective regulatory decision making of various committees. This article elaborates the maintenance of records and information at BSC

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.

TITAN Legal Weight Truck cask preliminary design report

International Nuclear Information System (INIS)

1990-04-01

The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of this document includes the LWT cask with fuel baskets, impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. The results of the tradeoff studies and evaluations that were performed during the preliminary design are presented in Appendix A to this report. 51 figs., 17 tabs

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

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.

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

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

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...

340 Waste Handling Facility interim safety basis

International Nuclear Information System (INIS)

Bendixsen, R.B.

1995-01-01

This document establishes the interim safety basis (ISB) for the 340 Waste Handling Facility (340 Facility). An ISB is a documented safety basis that provides a justification for the continued operation of the facility until an upgraded final safety analysis report is prepared that complies with US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports. The ISB for the 340 Facility documents the current design and operation of the facility. The 340 Facility ISB (ISB-003) is based on a facility walkdown and review of the design and operation of the facility, as described in the existing safety documentation. The safety documents reviewed, to develop ISB-003, include the following: OSD-SW-153-0001, Operating Specification Document for the 340 Waste Handling Facility (WHC 1990); OSR-SW-152-00003, Operating Limits for the 340 Waste Handling Facility (WHC 1989); SD-RE-SAP-013, Safety Analysis Report for Packaging, Railroad Liquid Waste Tank Cars (Mercado 1993); SD-WM-TM-001, Safety Assessment Document for the 340 Waste Handling Facility (Berneski 1994a); SD-WM-SEL-016, 340 Facility Safety Equipment List (Berneski 1992); and 340 Complex Fire Hazard Analysis, Draft (Hughes Assoc. Inc. 1994)

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

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.

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)

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)

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

New set of Chemical Safety rules

CERN Multimedia

HSE Unit

2011-01-01

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

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

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

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...

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.

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

Mixing of incompatible materials in waste tanks technical basis document

International Nuclear Information System (INIS)

SANDGREN, K.R.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials 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 (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the FR-equency 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

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

Health and Safety Work Plan for Sampling Colloids in Waste Area Grouping 5 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

Marsh, J.D.; McCarthy, J.F.

1994-01-01

This Work Plan/Site Safety and Health Plan (SSHP) and the attached work plan are for the performance of the colloid project at WAG 5. The work will be conducted by the Oak Ridge National Laboratory (ORNL) Environmental Sciences Division (ESD) and associated ORNL environmental, safety, and health support groups. The purpose of this document is to establish health and safety guidelines to be followed by all personnel involved in conducting work for this project. The levels of protection and the procedures specified in this plan are based on the best information available from historical data and preliminary evaluations of the area. Therefore, these recommendations represent the minimum health and safety requirements to be observed by all personnel engaged in this project

Health and Safety Work Plan for Sampling Colloids in Waste Area Grouping 5 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

Energy Technology Data Exchange (ETDEWEB)

Marsh, J.D.; McCarthy, J.F.

1994-01-01

This Work Plan/Site Safety and Health Plan (SSHP) and the attached work plan are for the performance of the colloid project at WAG 5. The work will be conducted by the Oak Ridge National Laboratory (ORNL) Environmental Sciences Division (ESD) and associated ORNL environmental, safety, and health support groups. The purpose of this document is to establish health and safety guidelines to be followed by all personnel involved in conducting work for this project. The levels of protection and the procedures specified in this plan are based on the best information available from historical data and preliminary evaluations of the area. Therefore, these recommendations represent the minimum health and safety requirements to be observed by all personnel engaged in this project.

Safety evaluation report of the Waste Isolation Pilot Plant safety analysis report: Contact-handled transuranic waste disposal operations

International Nuclear Information System (INIS)

1997-02-01

DOE 5480.23, Nuclear Safety Analysis Reports, requires that the US Department of Energy conduct an independent, defensible, review in order to approve a Safety Analysis Report (SAR). That review and the SAR approval basis is documented in this formal Safety Evaluation Report (SER). This SER documents the DOE's review of the Waste Isolation Pilot Plant SAR and provides the Carlsbad Area Office Manager, the WIPP SAR approval authority, with the basis for approving the safety document. It concludes that the safety basis documented in the WIPP SAR is comprehensive, correct, and commensurate with hazards associated with planned waste disposal operations

Development of a preliminary decommissioning plan of the reactor IPEN/MB-01

International Nuclear Information System (INIS)

Vivas, Ary de Souza; Carneiro, Alvaro Luiz Guimaraes

2013-01-01

Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend, through publications, guidelines for the conduct of activities both for decommissioning nuclear power plants and for research reactors, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors, having only a standard applied to decommissioning power plants which was published in November 2012. The Nuclear and Energy Research Institute (IPEN) has two research reactors one being the reactor IPEN/MB-01. The aim of this work is to develop a preliminary plan for decommissioning of nuclear reactor research, considering the technical documentation of the system (RAS-Safety Analysis Report), the existing rules of CNEN (National Nuclear Energy Commission), as well as regulatory instructions and recommendations of the IAEA. The preliminary decommissioning plan consists of the presentation of actions and steps required as well as the strategies to be adopted for the shutdown of the facility under the technical and administrative, seeking the safety, health workers and the general public, minimizing environmental impacts. (author)

NASA Hybrid Wing Aircraft Aeroacoustic Test Documentation Report

Science.gov (United States)

Heath, Stephanie L.; Brooks, Thomas F.; Hutcheson, Florence V.; Doty, Michael J.; Bahr, Christopher J.; Hoad, Danny; Becker, Lawrence; Humphreys, William M.; Burley, Casey L.; Stead, Dan;

EPR safety. Consideration of the internal and external hazards in the safety studies

International Nuclear Information System (INIS)

Gueguin, H.

2008-04-01

The author presents the main points of the Preliminary Safety Report of EDF on the EPR reactor safety. It concerns the considerations of the internal (fire, flood, explosions, pipes failures) and external (earthquakes, airplane falls, explosions, exceptional natural disasters, extreme meteorological conditions) damages. It presents how the safety report takes into account the aggression. (A.L.B.)

Safety practice and regulations in different IGORR member countries

International Nuclear Information System (INIS)

Hickman, C.; Minguet, J.L.; Arnould, F.

1999-01-01

In the suggestions of the 1996 IGORR 5 conference, Technicatome proposed 'Comparing Regulations for Research Reactors in Participating Countries'. The aim was to enhance and facilitate the dissemination of pertinent information amongst potential utilities of operational research reactors. A questionnaire on the following topics was subsequently sent out to IGORR 5 participants : Procedures for Research Reactors and Associated Equipment, Safety Analysis, Safety Related Components, Radiation Protection and Management of Nuclear materials. The objective of the present paper is to identify major trends, similarities and differences in the approaches adopted by different countries. Its scope has been limited to: Licensing and Regulatory approach; Operating and Safety documents; Safety Analysis; Radiological Safety; Management of Nuclear Materials. The investigations carried out indicate that to a large extent international recommendations (IAEA, ICPR,..) are being followed and that there is a general tendency to integrate them into national legislation and regulations. Although Safety Culture varies from one country to another an overall general consensus exists on the basic approach to safety inasmuch as: different countries have their own legally defined Safety Authorities, a Preliminary Safety Report is required before a research reactor can be built, and a final Safety Report before the core can be loaded with nuclear fuel and the reactor made critical; these documents must be accepted by the Safety Authorities concerned; a combination of defense-in-depth strategy (deterministic approach) and probabilistic analysis is applied; three or more safety classes are used to categorize systems and components; the single failure criterion is taken into consideration for systems and components having safety functions; both Operating Basis and Safety Shutdown type earthquakes are considered; the crashing of an aircraft onto a research reactor is taken into consideration

49 CFR 591.6 - Documents accompanying declarations.

Science.gov (United States)

2010-10-01

... SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) IMPORTATION OF VEHICLES AND EQUIPMENT SUBJECT TO FEDERAL SAFETY, BUMPER AND THEFT PREVENTION STANDARDS § 591.6 Documents accompanying... be accompanied by a statement substantiating that the vehicle was not manufactured for use on the...

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

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…

Guide for the realization of Design Base Documents (DBD)

International Nuclear Information System (INIS)

Roca Mallofre, G. la

2010-01-01

Guide for improving the consistency and quality content of the Design Base Documents. It's a short description of how to carry out and complete these Documents but focusing on those aspects that can be more confusing and harder to interpret. This guide aims to clarify the term Design Base distinguishing between production and safety, and it focuses on safety Design Base Documents and their values and references. It also emphasizes the difference between the support system and the interface system when there is a functional connection between different systems.

Hanford Waste Vitrification Plant: Preliminary description of waste form and canister

International Nuclear Information System (INIS)

Mitchell, D.E.

1986-01-01

In July 1985, the US Department of Energy's Office of Civilian Radioactive Waste Management established the Waste Acceptance Process as the means by which defense high-level waste producers, such as the Hanford Waste Vitrification Plant, will develop waste acceptance requirements with the candidate geologic repositories. A complete description of the Waste Acceptance Process is contained in the Preliminary Hanford Waste Vitrification Plant Waste Form Qualification Plan. The Waste Acceptance Process defines three documents that high-level waste producers must prepare as a part of the process of assuming that a high-level waste product will be acceptable for disposal in a geologic repository. These documents are the Description of Waste Form and Canister, Waste Compliance Plan, and Waste Qualification Report. This document is the Hanford Waste Vitrification Plant Preliminary Description of Waste Form and Canister for disposal of Neutralized Current Acid Waste. The Waste Acceptance Specifications for the Hanford Waste Vitrification Plant have not yet been developed, therefore, this document has been structured to corresponds to the Waste Acceptance Preliminary Specifications for the Defense Waste Processing Facility High-Level Waste Form. Not all of the information required by these specifications is appropriate for inclusion in this Preliminary Description of Waste Form and Canister. Rather, this description is limited to information that describes the physical and chemical characteristics of the expected high-level waste form. The content of the document covers three major areas: waste form characteristics, canister characteristics, and canistered waste form characteristics. This information will be used by the candidate geologic repository projects as the basis for preliminary repository design activities and waste form testing. Periodic revisions are expected as the Waste Acceptance Process progresses

Preliminary Development of Regulatory PSA Models for SFR

International Nuclear Information System (INIS)

Choi, Yong Won; Shin, Andong; Bae, Moohoon; Suh, Namduk; Lee, Yong Suk

2013-01-01

Well developed PRA methodology exists for LWR (Light Water Reactor) and PHWR (Pressurized Heavy Water Reactor). Since KAERI is developing a prototype SFR targeting to apply for a license by 2017, KINS needs to have a PRA models to assess the safety of this prototype reactor. The purpose of this study is to develop the regulatory PSA models for the independent verification of the SFR safety. Since the design of the prototype SFR is not mature yet, we have tried to develop the preliminary models based on the design data of KAERI's previous SFR design. In this study, the preliminary initiating events of level 1 internal event for SFR were selected through reviews of existing PRA (LWR, PRISM, ASTRID and KALIMER-600) models. Then, the event tree for each selected initiating event was developed. The regulatory PRA models of SFR developed are preliminary in a sense, because the prototype SFR design is not mature and provided yet. Still it might be utilized for the forthcoming licensing review in assessing the risk of safety issues and the configuration control of the design

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.

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

Safety report content and development for test loop facility on MARIA reactor

International Nuclear Information System (INIS)

Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

1982-01-01

A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

Community Documentation Centre on Industrial Risk. Bulletin no. 8

International Nuclear Information System (INIS)

Masera, M.; Rasmussen, K.

1993-01-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 4

International Nuclear Information System (INIS)

Gow, H.B.F.

1991-01-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 10

International Nuclear Information System (INIS)

Perschke, A.; Kirchsteiger, C.

1996-01-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 5

International Nuclear Information System (INIS)

Gow, H.B.F.

1991-11-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 7

International Nuclear Information System (INIS)

Gow, H.B.F.; Carditello, I.

1993-04-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 9

International Nuclear Information System (INIS)

Perschke, A.

1995-01-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 6

International Nuclear Information System (INIS)

Gow, H.B.F.

1992-06-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

Community Documentation Centre on Industrial Risk. Bulletin no. 11

International Nuclear Information System (INIS)

Perschke, A.; Kirchsteiger, C.; Carnevali, C.

1997-01-01

The Directorate-General for Environment, Nuclear Safety and Civil Protection of the Commission of the European Communities is responsible for the effective and harmonized implementation of the Directive 82/501/EEC on the major-accident hazards of certain industrial activities. To this end, the Commission, in collaboration with the Committee of Competent Authorities responsible for the implementation of this Directive in the twelve Member States, carries out a whole range of activities. One of the most essential areas for action identified was the need for a systematic diffusion of information concerning the practical implementation of the Directive in the Member States, including the technical rules and guidelines applied, the safety practices and the lessons learnt from major accidents. Therefore, the Commission decided to set up a Community Documentation Centre on Industrial Risks (CDCIR). This Documentation Centre is run by the European Commission, Joint Research Centre, Institute for Systems Engineering and Informatics (ISEI), at Ispra, Italy, among its support activities on the implementation of the Directive. The Documentation Centre will collect, classify and review technical rules, guidelines and documents concerning the requirements of the Directive, as well as the safety of industrial installations produced by governments, administrative, scientific or technical bodies, national or international organizations and industrial or professional associations. Documents on major accidents in the form of reports, videotapes will also be collected and reviewed. The Centre is accessible to interested visitors, documents which are not covered by copyright and are not restricted can be obtained from the Documentation Centre on request. Periodical volumes which will feature the inventory, including abstracts, of the collected material will be published and made available to all interested parties. The Centre will also publish documents devoted to compare existing

A new era of safety at CERN

CERN Multimedia

CERN Bulletin

2014-01-01

CERN is modernising its safety policy and organisational structure in matters of Safety with the introduction of new reference documents that have come into force on 29 September. These texts adapt the Organization’s safety policy to take account of how the Laboratory has evolved and to include best practice in Safety matters.   Safety is a priority at CERN, so it’s no coincidence that the Organization’s anniversary has been chosen as the time to launch a modernised approach to its Safety policy and how Safety matters are organised. On the day of CERN’s 60th anniversary, the SAPOCO 42 document, which covered both policy and organisational aspects, was replaced by a more concise general policy statement. The organisational structure and responsibilities in matters of Safety are now set out in a Safety Regulation, that is supplemented by subsidiary documents. Together these documents will replace the corresponding parts of the former SAPOCO 42 as well as Saf...

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.

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

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 .

Engineering design guidelines for nuclear criticality safety

International Nuclear Information System (INIS)

Waltz, W.R.

1988-08-01

This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

49 CFR 237.155 - Documents and records.

Science.gov (United States)

2010-10-01

..., DEPARTMENT OF TRANSPORTATION BRIDGE SAFETY STANDARDS Documentation, Records, and Audits of Bridge Management Programs § 237.155 Documents and records. Each track owner required to implement a bridge management... protected by a security system that incorporates a user identity and password, or a comparable method, to...

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.

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.

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

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.

Safety first. Status reports on the IAEA's safety standards

International Nuclear Information System (INIS)

Webb, G.; Karbassioun, A.; Linsley, G.; Rawl, R.

1998-01-01

Documents in the IAEA's Safety Standards Series known as RASS (Radiation Safety Standards) are produced to develop an internally consistent set of regulatory-style publications that reflects an international consensus on the principles of radiation protection and safety and their application through regulation. In this article are briefly presented the Agency's programmes on Nuclear Safety Standards (NUSS), Radioactive Waste Safety Standards (RADWASS), and Safe Transport of Radioactive Materials

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

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.

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

Nuclear criticality safety department training implementation

International Nuclear Information System (INIS)

Carroll, K.J.; Taylor, R.G.; Worley, C.A.

1996-01-01

The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. The NCSD Qualification Program is described in Y/DD-694, Qualification Program, Nuclear Criticality Safety Department This document provides a listing of the roles and responsibilities of NCSD personnel with respect to training and details of the Training Management System (TMS) programs, Mentoring Checklists and Checksheets, as well as other documentation utilized to implement the program. This document supersedes Y/DD-696, Revision 2, dated 3/27/96, Training Implementation, Nuclear Criticality Safety Department. There are no backfit requirements associated with revisions to this document

Automating the Generation of Heterogeneous Aviation Safety Cases

Science.gov (United States)

Denney, Ewen W.; Pai, Ganesh J.; Pohl, Josef M.

2012-01-01

A safety case is a structured argument, supported by a body of evidence, which provides a convincing and valid justification that a system is acceptably safe for a given application in a given operating environment. This report describes the development of a fragment of a preliminary safety case for the Swift Unmanned Aircraft System. The construction of the safety case fragment consists of two parts: a manually constructed system-level case, and an automatically constructed lower-level case, generated from formal proof of safety-relevant correctness properties. We provide a detailed discussion of the safety considerations for the target system, emphasizing the heterogeneity of sources of safety-relevant information, and use a hazard analysis to derive safety requirements, including formal requirements. We evaluate the safety case using three classes of metrics for measuring degrees of coverage, automation, and understandability. We then present our preliminary conclusions and make suggestions for future work.

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

Cold Vacuum Drying facility design basis accident analysis documentation

International Nuclear Information System (INIS)

CROWE, R.D.

2000-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility 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 FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

Cold Vacuum Drying facility design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

2000-08-08

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility 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 FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

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...

HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

Energy Technology Data Exchange (ETDEWEB)

EVANS, C B

2004-12-21

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

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.

Tank farms criticality safety manual

International Nuclear Information System (INIS)

FORT, L.A.

2003-01-01

This document defines the Tank Farms Contractor (TFC) criticality safety program, as required by Title 10 Code of Federal Regulations (CFR-), Subpart 830.204(b)(6), ''Documented Safety Analysis'' (10 CFR- 830.204 (b)(6)), and US Department of Energy (DOE) 0 420.1A, Facility Safety, Section 4.3, ''Criticality Safety.'' In addition, this document contains certain best management practices, adopted by TFC management based on successful Hanford Site facility practices. Requirements in this manual are based on the contractor requirements document (CRD) found in Attachment 2 of DOE 0 420.1A, Section 4.3, ''Nuclear Criticality Safety,'' and the cited revisions of applicable standards published jointly by the American National Standards Institute (ANSI) and the American Nuclear Society (ANS) as listed in Appendix A. As an informational device, requirements directly imposed by the CRD or ANSI/ANS Standards are shown in boldface. Requirements developed as best management practices through experience and maintained consistent with Hanford Site practice are shown in italics. Recommendations and explanatory material are provided in plain type

Hanford surplus facilities hazards identification document. Revision 2

International Nuclear Information System (INIS)

Egge, R.G.

1996-02-01

This document provides general safety information needed by personnel who enter and work in surplus facilities managed by Bechtel Hanford, Inc. (BHI). 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. Questions concerning the currency of this information should be directed to the building administrator (as listed in BHI-FS-01, Field Support Administration, Section 1.1, ''Access Control for ERC Surplus Facilities'')

Safety at Work : Research Methodology

NARCIS (Netherlands)

Beurden, van K. (Karin); Boer, de J. (Johannes); Brinks, G. (Ger); Goering-Zaburnenko, T. (Tatiana); Houten, van Y. (Ynze); Teeuw, W. (Wouter)

2012-01-01

In this document, we provide the methodological background for the Safety atWork project. This document combines several project deliverables as defined inthe overall project plan: validation techniques and methods (D5.1.1), performanceindicators for safety at work (D5.1.2), personal protection

International consensus on safety principles

International Nuclear Information System (INIS)

Warnecke, E.

1993-01-01

The International Atomic Energy Agency (IAEA) has been regularly requested by its Member States to provide evidence that radioactive waste can be managed safely and to help demonstrate a harmonization of approach at the international level by providing safety documents. In response, IAEA established a special series of safety documents devoted to radioactive waste management. These documents will be elaborated within the Radioactive Waste Safety Standards (RADWASS) programme [1,2] which covers all aspects of radioactive waste management. The RADWASS programme develops a series of international consensus documents on all parts of the safe management of radioactive waste, including disposal. The purpose of the RADWASS programme is to (i) document existing international consensus in the approaches and methodologies for safe radioactive waste management, (ii) create a mechanism to establish consensus where it does not exist and (iii) provide Member States with a comprehensive series of internationally agreed upon documents to complement national standards and criteria. This paper describes the RADWASS programme, and covers the structure, implementation plans and status of documents under preparation

TITAN Legal Weight Truck cask preliminary design report

International Nuclear Information System (INIS)

1990-04-01

The Preliminary Design of the TITAN Legal Weight Truck (LWT) Cask System and Ancillary Equipment is presented in this document. The scope of the document includes the LWT cask with fuel baskets; impact limiters, and lifting and tiedown features; the cask support system for transportation; intermodal transfer skid; personnel barrier; and cask lifting yoke assembly. 75 figs., 48 tabs

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.

HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

International Nuclear Information System (INIS)

EVANS, C.B.

2004-01-01

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

Preliminary hazards analysis -- vitrification process

International Nuclear Information System (INIS)

Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P.

1994-06-01

This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility's construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment

Preliminary hazards analysis -- vitrification process

Energy Technology Data Exchange (ETDEWEB)

Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P. [Science Applications International Corp., Pleasanton, CA (United States)

1994-06-01

This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility`s construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment.

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.)

Safety equipment list for the 241-SY-101 RAPID mitigation project

Energy Technology Data Exchange (ETDEWEB)

MORRIS, K.L.

1999-06-29

This document provides the safety classification for the safety (safety class and safety RAPID Mitigation Project. This document is being issued as the project SEL until the supporting authorization basis documentation, this document will be superseded by the TWRS SEL (LMHC 1999), documentation istlralized. Upon implementation of the authorization basis significant) structures, systems, and components (SSCS) associated with the 241-SY-1O1 which will be updated to include the information contained herein.

Safety equipment list for the 241-SY-101 RAPID mitigation project

International Nuclear Information System (INIS)

Morris, K.L.

1999-01-01

This document provides the safety classification for the safety (safety class and safety RAPID Mitigation Project. This document is being issued as the project SEL until the supporting authorization basis documentation, this document will be superseded by the TWRS SEL (LMHC 1999), documentation istlralized. Upon implementation of the authorization basis significant) structures, systems, and components (SSCS) associated with the 241-SY-1O1 which will be updated to include the information contained herein

Safety indicators: an efficient tool for a better safety

International Nuclear Information System (INIS)

Aufort, P.; Lars, R.

1993-01-01

Safety indicators based on the examination of the Operating Technical Specifications have been defined with the aim of following the in-operation safety level of French nuclear power plants. These safety indicators are operation feedback tools which permit the a posteriori justification and the adjustment of actual procedures. They would allow detection of an abnormal unavailability occurrence rate or a situation revealing a potential safety problem. So, data acquisition, processing, analysis and display software allowing trend analysis of these indicators has been developed so far as: a reflexion tool for the power plant operators about the safety instructions and the adjustment of preventive maintenance, and a help for decision making at a national level for the examination and the improvement of Operating Technical Specifications. This paper presents the objectives of these safety indicators, the processing tool associated, the preliminary results obtained and more elaborate processing of these indicators. These safety indicators may be very useful in framing probabilistic safety assessments. (author)

75 FR 50733 - Preliminary Theft Data; Motor Vehicle Theft Prevention Standard

Science.gov (United States)

2010-08-17

... [Docket No. NHTSA 2010-0098] Preliminary Theft Data; Motor Vehicle Theft Prevention Standard AGENCY... of preliminary theft data; request for comments. SUMMARY: This document requests comments on data about passenger motor vehicle thefts that occurred in calendar year (CY) 2008 including theft rates for...

Integrated Management System, Configuration and Document Control for Research Reactors

International Nuclear Information System (INIS)

Steynberg, B.J.; Bruyn, J.F. du

2017-01-01

An integrated management system is a single management framework establishing all the processes necessary for the organisation to address all its goals and objectives. Very often only quality, environment and health & safety goals are included when referred to an integrated management system. However, within the research reactor environment such system should include goals pertinent to economic, environmental, health, operational, quality, safeguards, safety, security, and social considerations. One of the important objectives of an integrated management is to create the environment for a healthy safety culture. Configuration management is a disciplined process that involves both management and technical direction to establish and document the design requirements and the physical configuration of the research reactor and to ensure that they remain consistent with each other and the documentation. Configuration is the combination of the physical, functional, and operational characteristics of the structures, systems, and components (SSCs) or parts of the research reactor, operation, or activity. The basic objectives and general principles of configuration management are the same for all research reactors. The objectives of configuration management are to: a) Establish consistency among design requirements, physical configuration, and documentation (including analyses, drawings, and procedures) for the research reactor; b) Maintain this consistency throughout the life of the research reactor, particularly as changes are being made; and c) Retain confidence in the safety of the research reactor. The key elements needed to manage the configuration of research reactors are design requirements, work control, change control, document control, and configuration management assessments. The objective of document control is to ensure that only the most recently approved versions of documents are used in the process of operating, maintaining, and modifying the research reactor

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

Engineering and Safety Partnership Enhances Safety of the Space Shuttle Program (SSP)

Science.gov (United States)

Duarte, Alberto

2007-01-01

Project Management must use the risk assessment documents (RADs) as tools to support their decision making process. Therefore, these documents have to be initiated, developed, and evolved parallel to the life of the project. Technical preparation and safety compliance of these documents require a great deal of resources. Updating these documents after-the-fact not only requires substantial increase in resources - Project Cost -, but this task is also not useful and perhaps an unnecessary expense. Hazard Reports (HRs), Failure Modes and Effects Analysis (FMEAs), Critical Item Lists (CILs), Risk Management process are, among others, within this category. A positive action resulting from a strong partnership between interested parties is one way to get these documents and related processes and requirements, released and updated in useful time. The Space Shuttle Program (SSP) at the Marshall Space Flight Center has implemented a process which is having positive results and gaining acceptance within the Agency. A hybrid Panel, with equal interest and responsibilities for the two larger organizations, Safety and Engineering, is the focal point of this process. Called the Marshall Safety and Engineering Review Panel (MSERP), its charter (Space Shuttle Program Directive 110 F, April 15, 2005), and its Operating Control Plan emphasizes the technical and safety responsibilities over the program risk documents: HRs; FMEA/CILs; Engineering Changes; anomalies/problem resolutions and corrective action implementations, and trend analysis. The MSERP has undertaken its responsibilities with objectivity, assertiveness, dedication, has operated with focus, and has shown significant results and promising perspectives. The MSERP has been deeply involved in propulsion systems and integration, real time technical issues and other relevant reviews, since its conception. These activities have transformed the propulsion MSERP in a truly participative and value added panel, making a

Safety culture in transport

International Nuclear Information System (INIS)

Decobert, V.

1998-01-01

'Safety culture' is a wording that appeared first in 1986, during the evaluation of what happened during the Tchernobyl accident. Safety culture is defined in the IAEA 75-INSAG-4 document as the characteristics and attitude which, in organizations and in men behaviours, make that questions related to safety of nuclear power plants benefits, in priority, of the attention that they need in function of their importance. The INSAG-4 document identifies three different elements necessary to the development of the safety culture: commitment of the policy makers, commitment of the managers of the industry, and commitment of individuals. This paper gives examples to show how safety culture is existing in the way Transnucleaire performs the activities in the field of transport of nuclear materials. (author)

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

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.

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...

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

International Nuclear Information System (INIS)

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

2001-01-01

This paper describes principles and approaches used by Gosatomnadzor of Russia in establishing safety goals. The link between safety goals and safety culture is demonstrated. The paper also contains information on nuclear regulatory activities in Russia. Regulatory documents of Gosatomnadzor of Russia do not provide precise definitions of safety goals as IAEA documents INSAG-3 or INSAG-12 do. However, overall activities of Gosatomnadzor of Russia are directed to the achievement of these safety goals, as Gosatomnadzor of Russia is a federal executive authority responsible for the regulation of nuclear and radiation safety in accordance with the Russian Federal Law 'On the Use of Nuclear Energy'. Thus, in the Statement of the Policy of the Russian Regulatory Authority, enacted in 1992, it was established that the overall activities of Gosatomnadzor of Russia are directed to the achievement of the main goal. This goal is to establish conditions that ensure that personnel, the public, and the environment are protected from unacceptable radiation and nonproliferation of nuclear materials. The practical application of such a method as given by the publication of Statements of Policy of Gosatomnadzor of Russia may be considered as a safety culture element. 'General Provisions of NPP Safety Ensuring' (OPB-88/ 97) is a regulatory document of the highest level in the hierarchy of regulatory documents of Gosatomnadzor of Russia. It establishes quantitative values of safety goals as do the foregoing IAEA documents. Thus, this regulatory document sets up the following: 1. The estimated total probability of severe accidents should not exceed 10 5 /reactor.yr. 2. The estimated probability of the worst possible radioactive release to the environment specified in the standards should not exceed 10 -7 /reactor.yr in the case of severe beyond-design-basis accidents. 3. The probability of a reactor vessel failure should not exceed 10 -7 /reactor.yr. The foregoing values are somehow

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

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)

Topical session proceedings of the 5. IGSC meeting on: observations regarding the safety case in recent safety assessment studies

International Nuclear Information System (INIS)

Hooper, Alan J.; Voinis, Sylvie; Van Luik, Abraham E.

2004-01-01

Within the NEA, the IGSC (Integration Group for the Safety Case) has, as an essential role, to develop common views on such key aspects of the safety case. Therefore, since the inauguration of the IGSC in 2000, four meetings were organised with topical sessions to explore various of these key aspects. This is a report on the fifth such topical session, held as part of the 5. plenary meeting of the IGSC. The session was attended by 36 participants, representing waste management organisations and regulatory authorities from 16 NEA member countries, the IAEA and the European Commission. The purpose of this topical session was to provide support to the finalising of the IGSC safety case brochure by getting a description of the safety case content of the IAEA Draft Safety Requirements document and by getting an overview of progress that could be observed from national organisations on developing their cases for system safety and/or developing the required methodologies. The objective was that the IGSC safety case brochure should be supportive of the IAEA/NEA document, and be reflective of the experience of the IGSC member programmes and organisations. The topical session was mainly aimed at exchanging information on: - The safety case related content of the proposed IAEA/NEA document (currently titled: 'IAEA Safety Standards Series, Geological Disposal of Radioactive Waste, Draft Safety Requirements (DS-154)'). - National programmes where safety assessments have recently been completed, e.g. ONDRAF/NIRAS, Nagra and Andra. - Feedback from international peer reviews, e.g. the Andra Dossier 2001 Argile, the Belgian SAFIR 2 report, the SR 97 report and the US-DOE Yucca Mountain TSPA. - The evolution of some national assessment methods and approaches e.g. SKB and Nagra. - The content of the draft IGSC safety case brochure entitled: 'The Nature and Purpose of the Post-closure Safety Case in Geological Disposal'. This document presents the various

Preliminary rock mechanics laboratory: Investigation plan

International Nuclear Information System (INIS)

Oschman, K.P.; Hummeldorf, R.G.; Hume, H.R.; Karakouzian, M.; Vakili, J.E.

1987-01-01

This document presents the rationale for rock mechanics laboratory testing (including the supporting analysis and numerical modeling) planned for the site characterization of a nuclear waste repository in salt. This plan first identifies what information is required for regulatory and design purposes, and then presents the rationale for the testing that satisfies the required information needs. A preliminary estimate of the minimum sampling requirements for rock laboratory testing during site characterization is also presented. Periodic revision of this document is planned

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

Setting clear expectations for safety basis development

International Nuclear Information System (INIS)

MORENO, M.R.

2003-01-01

DOE-RL has set clear expectations for a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (10 CFR 830, Nuclear Safety Rule) which will ensure long-term benefit to Hanford. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development resulting in a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was issued to standardized methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was issued for the evaluation of radiological consequences for accident scenarios often postulated for Hanford. A standard Site Documented Safety Analysis (DSA) detailing the safety management programs was issued for use as a means of compliance with a majority of 3009 Standard chapters. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. As a result of setting expectations and providing safety analysis tools, the four Hanford Site waste management nuclear facilities were able to integrate into one Master Waste Management Documented Safety Analysis (WM-DSA)

Improved safety at CERN

CERN Multimedia

2006-01-01

As announced in Weekly Bulletin No. 43/2006, a new approach to the implementation of Safety at CERN has been decided, which required taking some managerial decisions. The guidelines of the new approach are described in the document 'New approach to Safety implementation at CERN', which also summarizes the main managerial decisions I have taken to strengthen compliance with the CERN Safety policy and Rules. To this end I have also reviewed the mandates of the Safety Commission and the Safety Policy Committee (SAPOCO). Some details of the document 'Safety Policy at CERN' (also known as SAPOCO42) have been modified accordingly; its essential principles, unchanged, remain the basis for the safety policy of the Organisation. I would also like to inform you that I have appointed Dr M. Bona as the new Head of the Safety Commission until 31.12.2008, and that I will proceed soon to the appointment of the members of the new Safety Policy Committee. All members of the personnel are deemed to have taken note of the d...

On-site transportation and handling of uranium-233 special nuclear material: Preliminary hazards and accident analysis. Final

International Nuclear Information System (INIS)

Solack, T.; West, D.; Ullman, D.; Coppock, G.; Cox, C.

1995-01-01

U-233 Special Nuclear Material (SNM) currently stored at the T-Building Storage Areas A and B must be transported to the SW/R Tritium Complex for repackaging. This SNM is in the form of oxide powder contained in glass jars which in turn are contained in heat sealed double polyethylene bags. These doubled-bagged glass jars have been primarily stored in structural steel casks and birdcages for approximately 20 years. The three casks, eight birdcages, and one pail/pressure vessel will be loaded onto a transport truck and moved over an eight day period. The Preliminary Hazards and Accident Analysis for the on-site transportation and handling of Uranium-233 Special Nuclear Material, documented herein, was performed in accordance with the format and content guidance of DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, dated July 1994, specifically Chapter Three, Hazard and Accident Analysis. The Preliminary Hazards Analysis involved detailed walkdowns of all areas of the U-233 SNM movement route, including the T-Building Storage Area A and B, T-Building truck tunnel, and the roadway route. Extensive discussions were held with operations personnel from the Nuclear Material Control Group, Nuclear Materials Accountability Group, EG and G Mound Security and the Material Handling Systems Transportation Group. Existing documentation related to the on-site transportation of hazardous materials, T-Building and SW/R Tritium Complex SARs, and emergency preparedness/response documentation were also reviewed and analyzed to identify and develop the complete spectrum of energy source hazards

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)

Safety Review related to Commercial Grade Digital Equipment in Safety System

International Nuclear Information System (INIS)

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

2013-01-01

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

The value of safety and safety as a value

NARCIS (Netherlands)

Ratilainen, H.; Salminen, S.; Zwetsloot, G.I.J.M.; Perttula, P.; Starren, A.; Steijn, W.; Pahkin, K.; Drupsteen, L.; Puro, V.; Räsänen, T.; Aaltonen, M.; Berkers, F.; Kalakoski, V.

2016-01-01

The research presented in this document analyzes how safety values are defined and used in practice, in particular by managers, and how they affect employers’ and employees’ decisions and behaviour at work. The work comprises three complementary activities: a literature review on the value of safety

Pesticide management approach towards protecting the safety and health of farmers in Southeast Asia.

Science.gov (United States)

Mohammad, Norsyazwani; Abidin, Emilia Zainal; How, Vivien; Praveena, Sarva Mangala; Hashim, Zailina

2018-06-27

It is estimated that pesticide production and use have increased continuously in the countries of Southeast Asia in recent years. Within the context of protecting the safety and health of workers in the agricultural sector, there is an existing gap in the implementation of the pesticide management framework because safety and health effects arising from occupational exposures continue to be reported. This study aims to provide narrative similarities, differences and weaknesses of the existing pesticide management system in Southeast Asian countries (Brunei, Cambodia, Indonesia, Laos, Malaysia, Myanmar, Philippines, Singapore, Thailand and Vietnam) within the context of occupational safety and health. This is preliminary traditional review study. Pesticide regulation and management at the country level were identified using web-based search engines such as Scopus, ScienceDirect, PubMed and Google. Book, reports, legislation document and other documents retrieved were also gathered from international organizations and specific websites of governmental agency in Southeast Asian countries. The scope of this review is only limited to literature written in English. In total, 44 review articles, reports and documents were gathered for this study. The approach of pesticide management in protecting safety and health in the agricultural setting were benchmarked according to the elements introduced by the United States Environmental Protection Agency, namely, (1) the protection of workers and (2) the practice of safety. All countries have assigned a local authority and government organization to manage and control pesticide use in the agricultural sector. The countries with the highest usage of pesticide are Thailand, Philippines and Malaysia. Most Southeast Asian countries have emphasized safety practice in the management of pesticide usage, but there were less emphasis on the element of protection of workers within the framework in Indonesia, Myanmar, Thailand and Vietnam

Reviewing industrial safety in nuclear power plants

International Nuclear Information System (INIS)

1990-02-01

This document contains guidance and reference materials for Operational Safety Review Team (OSART) experts, in addition to the OSART Guidelines (TECDOC-449), for use in the review of industrial safety activities at nuclear power plants. It sets out objectives for an excellent industrial safety programme, and suggests investigations which should be made in evaluating industrial safety programmes. The attributes of an excellent industrial safety programme are listed as examples for comparison. Practical hints for reviewing industrial safety are discussed, so that the necessary information can be obtained effectively through a review of documents and records, discussions with counterparts, and field observations. There are several annexes. These deal with major features of industrial safety programmes such as safety committees, reporting and investigation systems and first aid and medical facilities. They include some examples which are considered commendable. The document should be taken into account not only when reviewing management, organization and administration but also in the review of related areas, such as maintenance and operations, so that all aspects of industrial safety in an operating nuclear power plant are covered

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

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.

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

Review of SKB's Code Documentation and Testing

International Nuclear Information System (INIS)

Hicks, T.W.

2005-01-01

SKB is in the process of developing the SR-Can safety assessment for a KBS 3 repository. The assessment will be based on quantitative analyses using a range of computational codes aimed at developing an understanding of how the repository system will evolve. Clear and comprehensive code documentation and testing will engender confidence in the results of the safety assessment calculations. This report presents the results of a review undertaken on behalf of SKI aimed at providing an understanding of how codes used in the SR 97 safety assessment and those planned for use in the SR-Can safety assessment have been documented and tested. Having identified the codes us ed by SKB, several codes were selected for review. Consideration was given to codes used directly in SKB's safety assessment calculations as well as to some of the less visible codes that are important in quantifying the different repository barrier safety functions. SKB's documentation and testing of the following codes were reviewed: COMP23 - a near-field radionuclide transport model developed by SKB for use in safety assessment calculations. FARF31 - a far-field radionuclide transport model developed by SKB for use in safety assessment calculations. PROPER - SKB's harness for executing probabilistic radionuclide transport calculations using COMP23 and FARF31. The integrated analytical radionuclide transport model that SKB has developed to run in parallel with COMP23 and FARF31. CONNECTFLOW - a discrete fracture network model/continuum model developed by Serco Assurance (based on the coupling of NAMMU and NAPSAC), which SKB is using to combine hydrogeological modelling on the site and regional scales in place of the HYDRASTAR code. DarcyTools - a discrete fracture network model coupled to a continuum model, recently developed by SKB for hydrogeological modelling, also in place of HYDRASTAR. ABAQUS - a finite element material model developed by ABAQUS, Inc, which is used by SKB to model repository buffer

Ferrocyanide Safety Program rationale for removing six tanks from the safety watch list

International Nuclear Information System (INIS)

Borsheim, G.L.

1993-09-01

This report documents an in-depth study of single-shell tanks containing ferrocyanide wastes. Topics include: safety assessments, tank histories, supportive documentation about interim stabilization and planned remedial activities

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

Safety assessment requirements for onsite transfers of radioactive material

International Nuclear Information System (INIS)

Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

1992-05-01

This document contains the requirements for developing a safety assessment document for an onsite package containing radioactive material. It also provides format and content guidance to establish uniformity in the safety assessment documentation and to ensure completeness of the information provided

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

A Preliminary Review of Fatigue Among Rail Staff

OpenAIRE

Jialin Fan; Andrew P. Smith

2018-01-01

Background: Fatigue is a severe problem in the rail industry, which may jeopardize train crew's health and safety. Nonetheless, a preliminary review of all empirical evidence for train crew fatigue is still lacking. The aim of the present paper is, therefore, to provide a preliminary description of occupational fatigue in the rail industry. This paper reviews the literature with the research question examining the risk factors associated with train crew fatigue, covering both papers published...

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

Safe Transport of Radioactive Material, International Regulations and its Supporting Documents

International Nuclear Information System (INIS)

El-Shinawy, R.M.K.

2005-01-01

Safe transport of radioactive material regulations issued by IAEA since 1961, provide standards for insuring a high level of safety of people,transport workers, property and environment against radiation, contamination and criticality hazards as well as thermal effects associated with the transport of the radioactive wastes and material. The history ,development, philosophy and scope of these international regulations were mentioned as well as the different supporting documents to the regulations for safe transport of radioactive material were identified.The first supporting document , namely TS - G-1.1 ( ST-2) ,Advisory material is also issued by the IAEA.It contains both the advisory and explanatory materials previously published in safety series No 7 and 37 and therefore TS-G-1.1 (ST-2) will supersede safety series No 7 and 37. The second supporting document namely TS-G-1.2 (ST-3), planning and preparing for emergency response to transport accidents involving radioactive material ,which will supersede safety series No 87. In addition to quality assurance (SS=113), compliance assurance (SS=112), the training manual and other

Safe Transport of Radioactive Material, International Regulations and its Supporting Documents

Energy Technology Data Exchange (ETDEWEB)

El-Shinawy, R M.K. [Radiation Protection Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

2005-04-01

Safe transport of radioactive material regulations issued by IAEA since 1961, provide standards for insuring a high level of safety of people,transport workers, property and environment against radiation, contamination and criticality hazards as well as thermal effects associated with the transport of the radioactive wastes and material. The history ,development, philosophy and scope of these international regulations were mentioned as well as the different supporting documents to the regulations for safe transport of radioactive material were identified.The first supporting document , namely TS - G-1.1 ( ST-2) ,Advisory material is also issued by the IAEA.It contains both the advisory and explanatory materials previously published in safety series No 7 and 37 and therefore TS-G-1.1 (ST-2) will supersede safety series No 7 and 37. The second supporting document namely TS-G-1.2 (ST-3), planning and preparing for emergency response to transport accidents involving radioactive material ,which will supersede safety series No 87. In addition to quality assurance (SS=113), compliance assurance (SS=112), the training manual and other.

Seismic safety of Paks nuclear power plant

International Nuclear Information System (INIS)

Katona, T.

1993-01-01

An extensive program is underway at Paks NPP for evaluation of the seismic safety and for development of the necessary safety increasing measures. This program includes the following five measures: investigation of methods, regulations and techniques utilized for reassessment of seismic safety of operating NPPs and promoting safety; investigation of earthquake hazards; development of concepts for creating the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept, and preliminary evaluation of the seismic safety

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.

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

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

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.)

Software Safety Risk in Legacy Safety-Critical Computer Systems

Science.gov (United States)

Hill, Janice L.; Baggs, Rhoda

2007-01-01

Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

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

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

Energy Technology Data Exchange (ETDEWEB)

E.N. Lindner

2004-12-03

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

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

International Nuclear Information System (INIS)

E.N. Lindner

2004-01-01

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

Early Childhood Safety Education: An Overview of Safety Curriculum in Outer Metropolitan, Regional and Rural NSW

Science.gov (United States)

Barr, Jennifer; Saltmarsh, Sue; Klopper, Christopher

2009-01-01

This article reports on preliminary findings from a 2008 survey and telephone interviews with 27 directors of early childhood education and care (ECEC) services located in regional and rural districts of the Australian state of New South Wales. Data from the study suggests that some areas of safety education--most notably road/traffic safety and…

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

Measures to strengthen international co-operation in nuclear, radiation and transport safety and waste management. Nuclear safety review for the year 2003

International Nuclear Information System (INIS)

2004-01-01

The Nuclear Safety Review for the Year 2003 presents an overview of the current issues and trends in nuclear, radiation, transport and radioactive waste safety during 2003. As in 2002 the overview is supported by more detailed Notes by the Secretariat: Safety Related Events and Issues Worldwide during 2003 (document 2004/Note 6), The Agency's Safety Standards: Activities during 2003 (document 2004/Note 7) and Providing for the Application of the Safety Standards (document 2004/Note 8). In January 2003, the Agency implemented an organization change and developed an integrated approach to reflect a broader assignment of nuclear safety and nuclear security and to better exploit synergy between them. The Office of Physical Protection and Material Security renamed to Office of Nuclear Security was transferred from the Department of Safeguards to the Department of Nuclear Safety, which became the Department of Nuclear Safety and Security to reflect the change. This Review provides information primarily on nuclear safety, and nuclear security will be addressed in a separate report

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

International Nuclear Information System (INIS)

1994-01-01

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

The Effects of Tabular-Based Content Extraction on Patent Document Clustering

Directory of Open Access Journals (Sweden)

Michael W. Berry

2012-10-01

Full Text Available Data can be represented in many different ways within a particular document or set of documents. Hence, attempts to automatically process the relationships between documents or determine the relevance of certain document objects can be problematic. In this study, we have developed software to automatically catalog objects contained in HTML files for patents granted by the United States Patent and Trademark Office (USPTO. Once these objects are recognized, the software creates metadata that assigns a data type to each document object. Such metadata can be easily processed and analyzed for subsequent text mining tasks. Specifically, document similarity and clustering techniques were applied to a subset of the USPTO document collection. Although our preliminary results demonstrate that tables and numerical data do not provide quantifiable value to a document’s content, the stage for future work in measuring the importance of document objects within a large corpus has been set.

Nuclear Criticality Safety Department Qualification Program

International Nuclear Information System (INIS)

Carroll, K.J.; Taylor, R.G.; Worley, C.A.

1996-01-01

The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of highly qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document defines the Qualification Program to address the NCSD technical and managerial qualification as required by the Y-1 2 Training Implementation Matrix (TIM). This Qualification Program is in compliance with DOE Order 5480.20A and applicable Lockheed Martin Energy Systems, Inc. (LMES) and Y-1 2 Plant procedures. It is implemented through a combination of WES plant-wide training courses and professional nuclear criticality safety training provided within the department. This document supersedes Y/DD-694, Revision 2, 2/27/96, Qualification Program, Nuclear Criticality Safety Department There are no backfit requirements associated with revisions to this document

Overview of the fundamental safety principles

International Nuclear Information System (INIS)

Chishinga, Milton Mulenga

2015-02-01

The primary objective of this work was to provide an overview of the International Atomic Energy (IAEA) document; 'Fundamental Safety principles, SF.1'. The document outlines ten (10) fundamental principles which provide the basis for an effective the radiation protection framework. The document is the topmost in the hierarchy of the IAEA Safety Standards Series. These principles are the foundation of the nuclear safety put stringent obligations on Parties under the Convention on Nuclear Safety. The fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. The fundamental Safety objective of protecting people individually and collectively and the environment has to be achieved without unduly limiting the operation of facilities or the conduct of activities that give rise to risks. The thematic areas covered are; responsibility for safety, role of government, leadership and management for safety, justification of facilities and activities, optimization of protection, limitation of risks to individuals, protection of present and future generations, prevention of accidents, emergency preparedness and response and protective actions to reduce existing or unregulated radiation risks. Appropriate recommendations have been provided for effective application of the principles by Governments, Regulatory Bodies and Operating Organizations of facilities and Nuclear Installations the give rise to radiation risks. (au)

B plant standards/requirements identification document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Maddox, B.S., Westinghouse Hanford

1996-07-29

This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES{ampersand}H) standards/requirements for the B Plant. This S/RID is applicable to the appropriate life cycle phases of design, construction,operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

Preliminary plan for decommissioning - repository for spent nuclear fuel

International Nuclear Information System (INIS)

Hallberg, Bengt; Tiberg, Liselotte

2010-06-01

The final disposal facility for spent nuclear fuel is part of the KBS-3 system, which also consists of a central facility for interim storage and encapsulation of the spent nuclear fuel and a transport system. The nuclear fuel repository will be a nuclear facility. Regulation SSMFS 2008:1 (Swedish Radiation Safety Authority's regulations on safety of nuclear facilities) requires that the licensee must have a current decommissioning plan throughout the facility lifecycle. Before the facility is constructed, a preliminary decommissioning plan should be reported to the Swedish Radiation Safety Authority. This document is a preliminary decommissioning plan, and submitted as an attachment to SKB's application for a license under the Nuclear Activities Act to construct, own and operate the facility. The final disposal facility for spent nuclear fuel consists of an above ground part and a below ground part and will be built near Forsmark and the final repository for radioactive operational waste, SFR. The parts above and below ground are connected by a ramp and several shafts, e.g. for ventilation. The below ground part consists of a central area, and several landfill sites. The latter form the repository area. The sealed below ground part constitutes the final repository. The decommissioning is taking place after the main operation has ended, that is, when all spent nuclear fuel has been deposited and the deposition tunnels have been backfilled and plugged. The decommissioning involves sealing of the remaining parts of the below ground part and demolition of above ground part. When decommissioning begins, there will be no contamination in the facility. The demolition is therefore performed as for a conventional plant. Demolition waste is sorted and recycled whenever possible or placed in landfill. Hazardous waste is managed in accordance with current regulations. A ground investigation is performed and is the basis for after-treatment of the site. The timetable for the

The IAEA safety standards

International Nuclear Information System (INIS)

Karbassioun, Ahmad

1995-01-01

During the development of the NUSS standards, wide consultation was carried out with all the Member States to obtain a consensus and the programme was supervised by a Senior Advisory Group consisting of senior safety experts from 13 countries. This group of senior regulators later became what is now known as the Nuclear Safety Standards Advisory Group (NUSSAG) and comprises of senior regulatory experts from 16 countries. The standards that were developed comprise of four types of documents: safety fundamentals; codes of practice; safety guides; and safety practices. The safety fundamentals set out the basic objectives, concepts and principles for nuclear safety in nuclear power plants. The codes of practice, are of a legislative nature, and establish the general objectives that must be fulfilled to ensure adequate nuclear power plant safety. They cover five areas: governmental organization; siting, design, operation and quality assurance. The safety guides, administrative in character, recommend procedures and acceptable technical solutions to implement the codes and guides by presenting further details gained from Member States, on the application and interpretation of individual concepts in the NUSS codes and guides. In total in the NUSS series there is currently one Fundamentals document, five Codes of Practice and fifty-six Safety Guides

Clinical decision support improves quality of telephone triage documentation--an analysis of triage documentation before and after computerized clinical decision support.

Science.gov (United States)

North, Frederick; Richards, Debra D; Bremseth, Kimberly A; Lee, Mary R; Cox, Debra L; Varkey, Prathibha; Stroebel, Robert J

2014-03-20

Clinical decision support (CDS) has been shown to be effective in improving medical safety and quality but there is little information on how telephone triage benefits from CDS. The aim of our study was to compare triage documentation quality associated with the use of a clinical decision support tool, ExpertRN©. We examined 50 triage documents before and after a CDS tool was used in nursing triage. To control for the effects of CDS training we had an additional control group of triage documents created by nurses who were trained in the CDS tool, but who did not use it in selected notes. The CDS intervention cohort of triage notes was compared to both the pre-CDS notes and the CDS trained (but not using CDS) cohort. Cohorts were compared using the documentation standards of the American Academy of Ambulatory Care Nursing (AAACN). We also compared triage note content (documentation of associated positive and negative features relating to the symptoms, self-care instructions, and warning signs to watch for), and documentation defects pertinent to triage safety. Three of five AAACN documentation standards were significantly improved with CDS. There was a mean of 36.7 symptom features documented in triage notes for the CDS group but only 10.7 symptom features in the pre-CDS cohort (p < 0.0001) and 10.2 for the cohort that was CDS-trained but not using CDS (p < 0.0001). The difference between the mean of 10.2 symptom features documented in the pre-CDS and the mean of 10.7 symptom features documented in the CDS-trained but not using was not statistically significant (p = 0.68). CDS significantly improves triage note documentation quality. CDS-aided triage notes had significantly more information about symptoms, warning signs and self-care. The changes in triage documentation appeared to be the result of the CDS alone and not due to any CDS training that came with the CDS intervention. Although this study shows that CDS can improve documentation, further study is needed

Preliminary Hazards Analysis Plasma Hearth Process

International Nuclear Information System (INIS)

Aycock, M.; Coordes, D.; Russell, J.; TenBrook, W.; Yimbo, P.

1993-11-01

This Preliminary Hazards Analysis (PHA) for the Plasma Hearth Process (PHP) follows the requirements of United States Department of Energy (DOE) Order 5480.23 (DOE, 1992a), DOE Order 5480.21 (DOE, 1991d), DOE Order 5480.22 (DOE, 1992c), DOE Order 5481.1B (DOE, 1986), and the guidance provided in DOE Standards DOE-STD-1027-92 (DOE, 1992b). Consideration is given to ft proposed regulations published as 10 CFR 830 (DOE, 1993) and DOE Safety Guide SG 830.110 (DOE, 1992b). The purpose of performing a PRA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PRA then is followed by a Preliminary Safety Analysis Report (PSAR) performed during Title I and II design. This PSAR then leads to performance of the Final Safety Analysis Report performed during construction, testing, and acceptance and completed before routine operation. Radiological assessments indicate that a PHP facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous material assessments indicate that a PHP facility will be a Low Hazard facility having no significant impacts either onsite or offsite to personnel and the environment

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.

Monitored Retrievable Storage System Requirements Document

International Nuclear Information System (INIS)

1994-03-01

This Monitored Retrievable Storage System Requirements Document (MRS-SRD) describes the functions to be performed and technical requirements for a Monitored Retrievable Storage (MRS) facility subelement and the On-Site Transfer and Storage (OSTS) subelement. The MRS facility subelement provides for temporary storage, at a Civilian Radioactive Waste Management System (CRWMS) operated site, of spent nuclear fuel (SNF) contained in an NRC-approved Multi-Purpose Canister (MPC) storage mode, or other NRC-approved storage modes. The OSTS subelement provides for transfer and storage, at Purchaser sites, of spent nuclear fuel (SNF) contained in MPCs. Both the MRS facility subelement and the OSTS subelement are in support of the CRWMS. The purpose of the MRS-SRD is to define the top-level requirements for the development of the MRS facility and the OSTS. These requirements include design, operation, and decommissioning requirements to the extent they impact on the physical development of the MRS facility and the OSTS. The document also presents an overall description of the MRS facility and the OSTS, their functions (derived by extending the functional analysis documented by the Physical System Requirements (PSR) Store Waste Document), their segments, and the requirements allocated to the segments. In addition, the top-level interface requirements of the MRS facility and the OSTS are included. As such, the MRS-SRD provides the technical baseline for the MRS Safety Analysis Report (SAR) design and the OSTS Safety Analysis Report design

High level waste storage tanks 242-A evaporator standards/requirement identification document

International Nuclear Information System (INIS)

Biebesheimer, E.

1996-01-01

This document, the Standards/Requirements Identification Document (S/RIDS) for the subject facility, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment. It lists those source documents from which requirements were extracted, and those requirements documents considered, but from which no requirements where taken. Documents considered as source documents included State and Federal Regulations, DOE Orders, and DOE Standards

Safety KPIs - Monitoring of safety performance

Directory of Open Access Journals (Sweden)

Andrej Lališ

2014-09-01

Full Text Available This paper aims to provide brief overview of aviation safety development focusing on modern trends represented by implementation of Safety Key Performance Indicators. Even though aviation is perceived as safe means of transport, it is still struggling with its complexity given by long-term growth and robustness which it has reached today. Thus nowadays safety issues are much more complex and harder to handle than ever before. We are more and more concerned about organizational factors and control mechanisms which have potential to further increase level of aviation safety. Within this paper we will not only introduce the concept of Key Performance Indicators in area of aviation safety as an efficient control mechanism, but also analyse available legislation and documentation. Finally we will propose complex set of indicators which could be applied to Czech Air Navigation Service Provider.

The Implementation of Cosine Similarity to Calculate Text Relevance between Two Documents

Science.gov (United States)

Gunawan, D.; Sembiring, C. A.; Budiman, M. A.

2018-03-01

Rapidly increasing number of web pages or documents leads to topic specific filtering in order to find web pages or documents efficiently. This is a preliminary research that uses cosine similarity to implement text relevance in order to find topic specific document. This research is divided into three parts. The first part is text-preprocessing. In this part, the punctuation in a document will be removed, then convert the document to lower case, implement stop word removal and then extracting the root word by using Porter Stemming algorithm. The second part is keywords weighting. Keyword weighting will be used by the next part, the text relevance calculation. Text relevance calculation will result the value between 0 and 1. The closer value to 1, then both documents are more related, vice versa.

[Patient safety in home care - A review of international recommendations].

Science.gov (United States)

Czakert, Judith; Lehmann, Yvonne; Ewers, Michael

2018-06-08

In recent years there has been a growing trend towards nursing care at home in general as well as towards intensive home care being provided by specialized home care services in Germany. However, resulting challenges for patient safety have rarely been considered. Against this background we aimed to explore whether international recommendations for patient safety in home care in general and in intensive home care in particular already exist and how they can stimulate further practice development in Germany. A review of online English documents containing recommendations for patient safety in intensive home care was conducted. Available documents were analyzed and compared in terms of their form and content. Overall, a small number of relevant documents could be identified. None of these documents exclusively refer to the intensive home care sector. Despite their differences, however, the analysis of four selected documents showed similarities, e. g., regarding specific topics of patient safety (communication, involvement of patients and their relatives, risk assessment, medication management, qualification). Furthermore, strengths and weaknesses of the documents became apparent: e. g., an explicit understanding of patient safety, a literature-based introduction to safety topics or an adaptation of the recommendations to the specific features of home care were occasionally lacking. This document analysis provides interesting input to the formal and content-related development of specific recommendations and to practice development in Germany to improve patient safety in home care. Copyright © 2018. Published by Elsevier GmbH.

Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

International Nuclear Information System (INIS)

Song, Tae Young

2007-01-01

At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas

Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Young [Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

2007-07-01

At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas.

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.

Cognitive Task Analysis of Business Jet Pilots' Weather Flying Behaviors: Preliminary Results

Science.gov (United States)

Latorella, Kara; Pliske, Rebecca; Hutton, Robert; Chrenka, Jason

2001-01-01

This report presents preliminary findings from a cognitive task analysis (CTA) of business aviation piloting. Results describe challenging weather-related aviation decisions and the information and cues used to support these decisions. Further, these results demonstrate the role of expertise in business aviation decision-making in weather flying, and how weather information is acquired and assessed for reliability. The challenging weather scenarios and novice errors identified in the results provide the basis for experimental scenarios and dependent measures to be used in future flight simulation evaluations of candidate aviation weather information systems. Finally, we analyzed these preliminary results to recommend design and training interventions to improve business aviation decision-making with weather information. The primary objective of this report is to present these preliminary findings and to document the extended CTA methodology used to elicit and represent expert business aviator decision-making with weather information. These preliminary findings will be augmented with results from additional subjects using this methodology. A summary of the complete results, absent the detailed treatment of methodology provided in this report, will be documented in a separate publication.

29 CFR 1921.17 - Service; copies of documents and pleadings.

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false Service; copies of documents and pleadings. 1921.17 Section 1921.17 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION... LONGSHOREMEN'S AND HARBOR WORKERS' COMPENSATION ACT Miscellaneous § 1921.17 Service; copies of documents and...

System Design and the Safety Basis

International Nuclear Information System (INIS)

Ellingson, Darrel

2008-01-01

The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

High-level waste tank farm set point document

International Nuclear Information System (INIS)

Anthony, J.A. III.

1995-01-01

Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope

High-level waste tank farm set point document

Energy Technology Data Exchange (ETDEWEB)

Anthony, J.A. III

1995-01-15

Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

Preliminary risk assessments of the small HTGR

International Nuclear Information System (INIS)

Everline, C.J.; Bellis, E.A.

1985-05-01

Preliminary investment and safety risk assessments were performed for a preconceptual design of a four-module 250-MW(t) side-by-side steel-vessel pebble bed HTGR plant. Broad event spectra were analyzed involving component damage resulting in unscheduled plant outages and fission product releases resulting in offsite doses. The preliminary assessment indicates at this stage of the design that two categories of events govern the investment risk envelope: primary coolant leaks which release some circulating and plate-out activity that contaminates the confinement and turbogenerator damage which involves extensive turbine blade failure. Primary coolant leaks are important contributors because associated cleanup and decontamination requirements result in longer outages that arise from other events with comparable frequencies. Turbogenerator damage is the salient low-frequency investment risk accident due to the relatively long outages being experienced in the industry. Thermal transients are unimportant investment risk contributors because pressurized core heatups cause little damage, and depressurized core heatups occur at negligible frequencies relative to the forced outage goal. These preliminary results demonstrate investment and safety risk goal compliance at this stage in the design process. Studies are continuing in order to provide valuable insights into risk-significant events to assure a balanced approach to meeting user and regulatory requirements

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

Management services, quality assurance, and safety

International Nuclear Information System (INIS)

Anon.

1987-01-01

Broad technical and administrative support for the programmatic research and development activities of the Fusion Energy Division is provided by the Management Services Section and by the division's quality assurance (QA) and safety programs. Support is provided through effective communication with division programmatic staff and through the coordination of resources from disciplines outside the division. The QA activity in the division emphasizes the development and documentation of a QA program that conforms to national standards, the review and approval of engineering documents, supplier surveillance, identification and documentation of nonconforming items, audits, and QA assessments/plans. The division's safety activities include a formal safety program, emergency planning activities, and environmental protection services. Efforts devoted to the removal of hazardous wastes from division facilities were expanded during 1986

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

French 900 MWe PWR PSA preliminary results

International Nuclear Information System (INIS)

Lanore, J.M.; Brisbois, J.

1988-10-01

A PSA is performed by the Safety Assessment Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the relative preliminary results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures

Preliminary waste acceptance requirements for the planned Konrad repository

International Nuclear Information System (INIS)

Warnecke, E.; Brennecke, P.

1987-01-01

The Physikalisch-Technische Bundesanstalt (PTB) has established Preliminary Waste Acceptance Requirements for the planned Konrad repository. These requirements were developed, in accordance with the Safety Criteria of the Reactor Safety Commission, with the help of a site specific safety assessment; they are under the reservation of the plan approval procedure, which is still in progress. In developing waste acceptance requirements, the PTB fulfills one of its duties as the institute responsible for waste disposal and gives guidelines for waste conditioning to waste producers and conditioners. (orig.) [de

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

EXPLOSION POTENTIAL ASSESSMENT OF HEAT EXCHANGER NETWORK AT THE PRELIMINARY DESIGN STAGE

Directory of Open Access Journals (Sweden)

MOHSIN PASHA

2016-07-01

Full Text Available The failure of Shell and Tube Heat Exchangers (STHE is being extensively observed in the chemical process industries. This failure can cause enormous production loss and have a potential of dangerous consequences such as an explosion, fire and toxic release scenarios. There is an urgent need for assessing the explosion potential of shell and tube heat exchanger at the preliminary design stage. In current work, inherent safety index based approach is used to resolve the highlighted issue. Inherent Safety Index for Shell and Tube Heat Exchanger (ISISTHE is a newly developed index for assessing the inherent safety level of a STHE at the preliminary design stage. This index is composed of preliminary design variables and integrated with the process design simulator (Aspen HYSYS. Process information can easily be transferred from process design simulator to MS Excel spreadsheet owing to this integration. This index could potentially facilitate the design engineer to analyse the worst heat exchanger in the heat exchanger network. Typical heat exchanger network of the steam reforming process is presented as a case study and the worst heat exchanger of this network has been identified. It is inferred from this analysis that shell and tube heat exchangers possess high operating pressure, corrected mean temperature difference (CMTD and flammability and reactive potential needs to be critically analysed at the preliminary design stage.

Safety barriers and safety functions a comparison of different applications

International Nuclear Information System (INIS)

Harms-Ringdahl, L.

1998-01-01

A study is being made with the focus on different theories and applications concerning 'safety barriers' and 'safety functions'. One aim is to compare the characteristics of different kinds of safely functions, which can be purpose, efficiency, reliability, weak points etc. A further aim is to summarize how the combination of different barriers are described and evaluated. Of special interest are applications from nuclear and chemical process safety. The study is based on a literature review, interviews and discussions. Some preliminary conclusions are made. For example, it appears to exist a need for better tools to support the design and evaluation of procedures. There are a great number of theoretical models describing safety functions. However, it still appears to be an interest in further development of models, which might give the basis for improved practical tools. (author)

76 FR 52945 - Chlorpyrifos Registration Review; Preliminary Human Health Risk Assessment; Extension of Comment...

Science.gov (United States)

2011-08-24

... ENVIRONMENTAL PROTECTION AGENCY [EPA-HQ-OPP-2008-0850; FRL-8886-6] Chlorpyrifos Registration... chlorpyrifos registration review; preliminary human health risk assessment. This document extends the comment... . SUPPLEMENTARY INFORMATION: This document extends the public comment period for the chlorpyrifos reregistration...

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...

Safety Design Criteria of Indian Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

2013-01-01

• Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

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.

Opinions on SKB's Safety Assessments SR 97 and SFL 3-5. A Review by SKI Consultants

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-12-01

The Swedish Nuclear Fuel and Waste Management Co. (SKB) has presented their safety assessment 'Deep repository for spent nuclear fuel, SR 97 - Post-closure safety'. SKB's report is part of the documentation that has been required by the Government before the start of site investigations. The Swedish Nuclear Power Inspectorate (SKI) is reviewing SR 97 according to earlier Government decisions. In its review work SKI has asked several consultants, that recently have been performing research work for SKI, to give their opinions on SR 97. SKI and the Swedish Radiation Protection Institute (SSI) have used these reports from the consultants as one complementary basis for the formulation of the SKI/SSI review report. This is a compilation of the reports from the different consultants, and therefore the different contributions vary in length, style and language. Included are also two consultant reports, giving comments on SKB's preliminary safety assessment for SFL 3-5 (deep repository for long-lived low- and intermediate-level waste). The 17 contributions have all been separately indexed.

Opinions on SKB's Safety Assessments SR 97 and SFL 3-5. A Review by SKI Consultants

International Nuclear Information System (INIS)

2000-12-01

The Swedish Nuclear Fuel and Waste Management Co. (SKB) has presented their safety assessment 'Deep repository for spent nuclear fuel, SR 97 - Post-closure safety'. SKB's report is part of the documentation that has been required by the Government before the start of site investigations. The Swedish Nuclear Power Inspectorate (SKI) is reviewing SR 97 according to earlier Government decisions. In its review work SKI has asked several consultants, that recently have been performing research work for SKI, to give their opinions on SR 97. SKI and the Swedish Radiation Protection Institute (SSI) have used these reports from the consultants as one complementary basis for the formulation of the SKI/SSI review report. This is a compilation of the reports from the different consultants, and therefore the different contributions vary in length, style and language. Included are also two consultant reports, giving comments on SKB's preliminary safety assessment for SFL 3-5 (deep repository for long-lived low- and intermediate-level waste). The 17 contributions have all been separately indexed

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)

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

Study on the KALIMER safety approach

International Nuclear Information System (INIS)

Kim, Eui Kwang; Han, Do Hee; Kim, Young Cheol.

1997-01-01

This study describes KALIMER's safety approach, how to establish the safety criteria and temperature limit, how to define safety evaluation events, and some safety research and development needs items. It is recommended that the KALIMER's approach to safety use seven levels of safety design and a defense-in-depth design approach with particular emphasis on inherent passive features. In order to establish as set DBEs for KALIMER safety evaluation, the procedure is explained how to define safety evaluation events. Final selection is to be determined later with the final establishment of design concepts. On the basis of preliminary studies and evaluation of the plant safety related areas, the KALIMER and PRISM have following three main difference that may require special research and development for KALIMER. (author). 7 refs., 6 tabs., 6 figs

Development of digital library system on regulatory documents for nuclear power plants

International Nuclear Information System (INIS)

Lee, K. H.; Kim, K. J.; Yoon, Y. H.; Kim, M. W.; Lee, J. I.

2001-01-01

The main objective of this study is to establish nuclear regulatory document retrieval system based on internet. With the advancement of internet and information processing technology, information management patterns are going through a new paradigm. Getting along the current of the time, it is general tendency to transfer paper-type documents into electronic-type documents through document scanning and indexing. This system consists of nuclear regulatory documents, nuclear safety documents, digital library, and information system with index and full text

Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-12-15

Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process.

Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

International Nuclear Information System (INIS)

Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won

2008-12-01

Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process

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.

Preliminary design review report for K Basin Dose Reduction Project

International Nuclear Information System (INIS)

Blackburn, L.D.

1996-01-01

The strategy for reducing radiation dose, originating from radionuclides absorbed in the K East Basin concrete, is to raise the pool water level to provide additional shielding. This report documents a preliminary design review conducted to ensure that design approaches for cleaning/coating basin walls and modifying other basin components were appropriate. The conclusion of this review was that design documents presently conclusion of this review was that design documents presently completed or in process of modification are and acceptable basis for proceeding to complete the design

Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide

International Nuclear Information System (INIS)

2010-01-01

The objective of this Safety Guide is to provide recommendations for meeting the IAEA safety requirements in performing or managing a level 2 probabilistic safety assessment (PSA) project for a nuclear power plant; thus it complements the Safety Guide on level 1 PSA. One of the aims of this Safety Guide is to promote a standard framework, standard terms and a standard set of documents for level 2 PSAs to facilitate regulatory and external peer review of their results. It describes all elements of the level 2 PSA that need to be carried out if the starting point is a fully comprehensive level 1 PSA. Contents: 1. Introduction; 2. PSA project management and organization; 3. Identification of design aspects important to severe accidents and acquisition of information; 4. Interface with level 1 PSA: Grouping of sequences; 5. Accident progression and containment analysis; 6. Source terms for severe accidents; 7. Documentation of the analysis: Presentation and interpretation of results; 8. Use and applications of the PSA; Annex I: Example of a typical schedule for a level 2 PSA; Annex II: Computer codes for simulation of severe accidents; Annex III: Sample outline of documentation for a level 2 PSA study.

Preliminary decommissioning plan of the reactor IPEN-MB01

International Nuclear Information System (INIS)

Vivas, Ary de Souza

2014-01-01

Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend a set of guidelines for the conduct of activities of nuclear facilities, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors. However, in March 2011 a study committee was formed with the main task facing the issues of decommissioning of nuclear installations in Brazil, culminating in Resolution 133 of November 8, 2012, a standard project that treat about the Decommissioning of nucleoelectric plants. O Instituto de Pesquisas Energeticas e Nucleares (IPEN) has two research reactors one being the reactor IPEN/MB-01. The purpose of this master dissertation is to develop a preliminary plan for decommissioning this research reactor, considering the technical documentation of the facility (RAS-Safety Analysis Report), the existing standards of CNEN (National Nuclear Energy Commission), as well as IAEA recommendations. In terms of procedures for decommissioning research reactors, this work was based on what is most modern in experiences, strategies and lessons learned performed and documented in IAEA publications covering techniques and technologies for decommissioning. Considering these technical knowledge and due to the peculiarities of the facility, was selected to immediate dismantling strategy, which corresponds to the start of decommissioning activities once the installation is switched off, dividing it into work sectors. As a resource for monitoring and project management of reactor decommissioning and maintenance of records, we developed a database using Microsoft Access 2007, which contain all the items and

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

Technical safety requirements (TSR) for waste receiving and processing (WRAP) facility

International Nuclear Information System (INIS)

Weidert, J.R.

1997-01-01

The scope of this TSR document is based on the WRAP Final Safety Analysis Report (HNF-SD-W026-SAR-002) and supporting documents. The administrative controls set forth in this TSR document are derived from the WRAP Final Safety Analysis Report

Assessing verticalization effects on urban safety perception

OpenAIRE

Lourenço, Ricardo Barros

2017-01-01

We describe an experiment with the modeling of urban verticalization effects on perceived safety scores as obtained with computer vision on Google Streetview data for New York City. Preliminary results suggests that for smaller buildings (between one and seven floors), perceived safety increases with building height, but that for high-rise buildings, perceived safety decreases with increased height. We also determined that while height contributing for this relation, other zonal aspects also ...

Guidance for the definition and application of probabilistic safety criteria

International Nuclear Information System (INIS)

Holmberg, J.-E.; Knochenhauer, M.

2011-05-01

The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

Guidance for the definition and application of probabilistic safety criteria

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-15

The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

Probabilistic safety criteria at the safety function/system level

International Nuclear Information System (INIS)

1989-09-01

A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs

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

Nuclear maintenance strategy and first steps for preliminary maintenance plan of the EU HCLL & HCPB Test Blanket Systems

Energy Technology Data Exchange (ETDEWEB)

Galabert, Jose, E-mail: jose.galabert@f4e.europa.eu [F4E Fusion for Energy, EU Domestic Agency, c/Josep Pla, 2. B3, 08019, Barcelona (Spain); Hopper, Dave [AMEC Foster Wheeler, Faraday Street, Birchwood Park, WA3 6GN (United Kingdom); Neviere, Jean-Cristophe [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, 13067, St. Paul Lez Durance Cedex (France); Nodwell, David [CCFE, Culham Science Centre, Abingdon, OX14 3DB, Oxfordshire (United Kingdom); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, 13067, St. Paul Lez Durance Cedex (France); Poitevin, Yves; Ricapito, Italo [F4E Fusion for Energy, EU Domestic Agency, c/Josep Pla, 2. B3, 08019, Barcelona (Spain); White, Gareth [AMEC Foster Wheeler, Faraday Street, Birchwood Park, WA3 6GN (United Kingdom)

2017-03-15

Highlights: • Nuclear maintenance strategy for the two European (EU) Test Blanket Systems (TBS): i/. Helium Cooled Lead Lithium (HCLL) and ii/. Helium Cooled Pebble Bed (HCPB). • Preliminary identification of maintenance tasks for most relevant components of the EU HCLL & HCPB TBS. • Preliminary feasibility analysis for hands-on maintenance tasks of some relevant components of the European Test Blanket Systems. • Design recommendations for enhancement of the European Test Blanket Systems maintainability. - Abstract: This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q{sub 2} Getter Beds, identifying some design recommendations for their sound maintainability.

Nuclear maintenance strategy and first steps for preliminary maintenance plan of the EU HCLL & HCPB Test Blanket Systems

International Nuclear Information System (INIS)

Galabert, Jose; Hopper, Dave; Neviere, Jean-Cristophe; Nodwell, David; Pascal, Romain; Poitevin, Yves; Ricapito, Italo; White, Gareth

2017-01-01

Highlights: • Nuclear maintenance strategy for the two European (EU) Test Blanket Systems (TBS): i/. Helium Cooled Lead Lithium (HCLL) and ii/. Helium Cooled Pebble Bed (HCPB). • Preliminary identification of maintenance tasks for most relevant components of the EU HCLL & HCPB TBS. • Preliminary feasibility analysis for hands-on maintenance tasks of some relevant components of the European Test Blanket Systems. • Design recommendations for enhancement of the European Test Blanket Systems maintainability. - Abstract: This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q_2 Getter Beds, identifying some design recommendations for their sound maintainability.

DOE standard: Firearms safety

International Nuclear Information System (INIS)

1996-02-01

Information in this document is applicable to all DOE facilities, elements, and contractors engaged in work that requires the use of firearms as provided by law or contract. The standard in this document provides principles and practices for implementing a safe and effective firearms safety program for protective forces and for non-security use of firearms. This document describes acceptable interpretations and methods for meeting Order requirements

DOE standard: Firearms safety

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

Information in this document is applicable to all DOE facilities, elements, and contractors engaged in work that requires the use of firearms as provided by law or contract. The standard in this document provides principles and practices for implementing a safe and effective firearms safety program for protective forces and for non-security use of firearms. This document describes acceptable interpretations and methods for meeting Order requirements.

Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 32, Coil assembly documentation. Volume 5

International Nuclear Information System (INIS)

Weber, C.M.

1995-01-01

This document is intended to address the contract requirement for providing coil assembly documentation, as required in the applicable Statement of Work: 'Provide preliminary procedures and preliminary design and supporting analysis of the equipment, fixtures, and hardware required to integrate and align the impregnated coil assemblies with the coil cases and intercoil structure. Each of the three major processes associated with the coil case and intercoil structure (ICS), TF Case Fabrication, Coil Preparation for Case Assembly are examined in detail. The specific requirements, processes, equipment, and technical concerns for each of these assembly processes is presented

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

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.

Safety of installations

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 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

Safety design of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

1984-01-01

Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

Westinghouse Hanford Company (WHC) standards/requirements identification document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Bennett, G.L.

1996-03-15

This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES&H) standards/requirements for Westinghouse Hanford Company Level Programs, where implementation and compliance is the responsibility of these organizations. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

Seismic safety of Paks nuclear power plant

International Nuclear Information System (INIS)

Katona, T.

1995-01-01

This paper contains an overview of the results concerning the following activities: investigation of methods, regulations and techniques for reassessment of seismic safety of operating NPPs and upgrading of safety; investigation of earthquake hazards; development of concept for creation of the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept and preliminary evaluation of the seismic safety. It is limited on investigation of dynamic features of building structures, the building dynamical experiments and experimental investigation of the equipment

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)

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...

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