Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of real component geometries (phase 1). Final report

International Nuclear Information System (INIS)

Nau, Andreas; Scholtes, B.

2014-01-01

-axial plane stress states were taken into account in this study. In the case of component's edges minor influences on the calculated stresses could be observed using standard calibration coefficients. Thus, a specific calibration is not considered necessary. The effects on shoulders could be assessed analogously. In the case of the claddings a remarkable influence could only be observed, if their thickness is below 2 mm. Essential benefits were reached through pre-experiments. As a consequence, the experimental execution was improved in terms of the material removal technique. The experimental data preparation was investigated in addition, which lead to the approach not to condition the experimental strains but rather the calculated stresses. This method is important for the experiments of phase 2 within this project. As a consequence of a successful project, the present knowledge base will be considerably improved and will be available for various engineering fields. Especially, the quantitative consideration of real residual stress states for optimized component designs will be feasible and finally the consequences of residual stresses an the component's safety, which are used in nuclear facilities, can be evaluated. The findings of the first numerical and theoretical research period (phase 1) at Kassel University are documented in this report.

Statistical considerations on safety analysis

International Nuclear Information System (INIS)

Pal, L.; Makai, M.

2004-01-01

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

The consideration of ecological safety in judicial practice-also on the ecological safety legislation

Institute of Scientific and Technical Information of China (English)

L(U) Zhongmei

2006-01-01

Ecological safety has been one of the hot issues of environmental law in recent years.The maintenance of ecological safety has become one of the legislative principles,as exemplified by the revision of the Law of Sand Prevention and Sand.Management and the Law against Solid Waste Environmental Pollution,and the relevant rules that will be established.However actual cases will still happen,whether the legislators have made the statutory law or not.While scholars and legislators are debating,the judges have to handle cases and render judgments.Through the analysis of a case,this article will discuss the feasibility for judges to make ecological safety considerations in the judicial process by applying the principle of good faith and will also discuss the legislative issues related to ecological safety.

User safety considerations in lithium thionyl chloride batteries

Energy Technology Data Exchange (ETDEWEB)

Johnson, L.J.; Willis, A.H.

1981-01-01

The mechanics of failure of lithium batteries with thionyl chloride as a catholyte are discussed, and methods available to provide user safety are outlined. Advantages of using the batteries are discussed, including a high cell voltage of 3.6 V, a high specific energy density greater than 200 W-hr per pound, excellent voltage stability over 95% of the discharge life of the cell, low cost materials, and a storage life up to 10 years. Safety problems such as explosion, venting, leaking, and short circuit overheating are also discussed. A fault matrix is presented to relate battery hazards to the possible use and disposal conditions, and to determine levels of safety for the user system. A test plan is also developed which includes mechanical, thermal, electrical and chemical considerations, and can be organized into categories convenient to various test facility capabilities.

User safety considerations in lithium thionyl chloride batteries

Science.gov (United States)

Johnson, L. J.; Willis, A. H.

The mechanics of failure of lithium batteries with thionyl chloride as a catholyte are discussed, and methods available to provide user safety are outlined. Advantages of using the batteries are discussed, including a high cell voltage of 3.6 V, a high specific energy density greater than 200 W-hr per pound, excellent voltage stability over 95% of the discharge life of the cell, low cost materials, and a storage life up to 10 years. Safety problems such as explosion, venting, leaking, and short circuit overheating are also discussed. A fault matrix is presented to relate battery hazards to the possible use and disposal conditions, and to determine levels of safety for the user system. A test plan is also developed which includes mechanical, thermal, electrical and chemical considerations, and can be organized into categories convenient to various test facility capabilities.

Criticality safety and facility design considerations

International Nuclear Information System (INIS)

Waltz, W.R.

1991-06-01

Operations with fissile material introduce the risk of a criticality accident that may be lethal to nearby personnel. In addition, concerns over criticality safety can result in substantial delays and shutdown of facility operations. For these reasons, it is clear that the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The emphasis of this report will be placed on engineering design considerations in the prevention of criticality. The discussion will not include other important aspects, such as the physics of calculating limits nor criticality alarm systems

76 FR 17808 - Final Vehicle Safety Rulemaking and Research Priority Plan 2011-2013

Science.gov (United States)

2011-03-31

... [Docket No. NHTSA-2009-0108] Final Vehicle Safety Rulemaking and Research Priority Plan 2011- 2013 AGENCY... availability. SUMMARY: This document announces the availability of the Final NHTSA Vehicle Safety and Fuel.... This Priority Plan is an update to the Final Vehicle Safety Rulemaking and Research Priority Plan 2009...

Safety considerations in the fast reactor fuel cycle

International Nuclear Information System (INIS)

Baker, A.R.; Burton, W.R.; Taylor, H.A.

1977-01-01

The fuel cycle safety problems for fast reactors, as compared with thermal reactors, are enhanced by the higher fissile content and heat rating of the fuel. Additionally recycling leads to the build up of substantial isotopes which contribute to the alpha and neutron hazards. The plutonium arisings in a nuclear power reactor programme extending into the next century are discussed. A requirement is to be able to return the product plutonium to a reactor about 9 months after the end of irradiation and it is anticipated that progress will be made slowly towards this fuel cycle, having regard to the necessity for maintaining safe and reliable operations. Consideration of the steps in the fuel cycle has indicated that it will be best to store the irradiated fuel on the reactor sites while I131 decays and decay heat falls before transporting and a suitable transport flask is being developed. Reprocessing development work is aimed at the key area of fuel breakdown, the inter-relation of the fuel characteristics on the dissolution of the plutonium and a solvent extract cycle leading to a product suitable for a co-located fabrication plant. Because of the high activity of recycled fuel it is considered that fabrication must move to a fully remote operation as is already the case for reprocessing, and a gel precipitation process producing a vibro compacted fuel is under development for this purpose. The waste streams from the processing plants must be minimised, processed for recovery of plutonium where applicable and then conditioned so that the final products released from the processing cycle are acceptable for ultimate disposal. The safety aspects reviewed cover protection of operators, containment of radioactive materials, criticality and regulation of discharges to the environment

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

SFR Safety Considerations

International Nuclear Information System (INIS)

Glatz, Jean-Paul

2012-01-01

Objectives of the Safety and Operation Project: • analysis and experiments that support approaches and assess performance of specific safety features, • development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and • valorisation of reactor operation, from experience and testing in operating SFR plants

HERBE final safety report; HERBE Finalni sigurnosni izvestaj

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-07-01

The Final safety report of HERBE system constructed at the RB reactor consists of 13 chapters, as follows. Chapter 0 includes a summary and the contents of the Final safety report, fundamental characteristics of the system and conclusion remarks, with the license agreement of the Safety Committee of the Boris Kidric Institute. Chapter 1 describes and analyzes the site of the HERBE system, including demography, topography, meteorology, hydrology, geology, seismicity, ecology. Chapter 3 covers technical characteristics of the system, Chapter 4 deals with safety analysis, Chapter 5 describes organisation of construction and preliminary operational testing of the system. Chapter 6 deals with organisation and program of test and regular operation, relevant procedures. Chapter 7 defines operational conditions and constraints, Chapter 8 and describe methods and means of radiation protection and radioactive materials management respectively. Chapter 10 contains a review of emergency plans, measures and procedures for nuclear accident protection. Chapters 11 and 12 are concerned with quality assurance program and physical protection of the HERBE system and related nuclear material.

Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

International Nuclear Information System (INIS)

Weidert, J.R.

1997-01-01

This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

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

Science.gov (United States)

2010-01-01

...; technical information in final safety analysis report. The application must contain a final safety analysis... 10 Energy 2 2010-01-01 2010-01-01 false Contents of applications; technical information in final safety analysis report. 52.157 Section 52.157 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES...

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

Science.gov (United States)

2010-01-01

...; technical information in final safety analysis report. (a) The application must contain a final safety... 10 Energy 2 2010-01-01 2010-01-01 false Contents of applications; technical information in final safety analysis report. 52.79 Section 52.79 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES...

Safety considerations for continuous hydrogen production test apparatus with capacity of 50 N-litter hydrogen per hour

International Nuclear Information System (INIS)

Onuki, Kaoru; Akino, Norio; Shimizu, Saburo; Nakajima, Hayato; Higashi, Shunichi; Kubo, Shinji

2001-03-01

Since the thermochemical hydrogen production Iodine-Sulfur process decomposes water into hydrogen and oxygen using toxic chemicals such as sulfuric acid, iodine and hydriodic acid, safety considerations are very important in its research and development. Therefore, before construction of continuous hydrogen production test apparatus with capacity of 50 N-litter hydrogen per hour, comprehensive safety considerations were carried out to examine the design and construction works of the test apparatus, and the experimental plans using the apparatus. Emphasis was given on the safety considerations on prevention of breakage of glasswares and presumable abnormalities, accidents and their countermeasures. This report summarizes the results of the considerations. (author)

SKB's safety case for a final repository license application

International Nuclear Information System (INIS)

Hedin, Allan; Andersson, Johan

2014-01-01

The safety assessment SR-Site is a main component in SKB's license application, submitted in March 2011, to construct and operate a final repository for spent nuclear fuel at Forsmark in the municipality of Oesthammar, Sweden. Its role in the application is to demonstrate long-term safety for a repository at Forsmark. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. The principal regulatory acceptance criterion, issued by the Swedish Radiation Safety Authority (SSM), requires that the annual risk of harmful effects after closure not exceed 10 -6 for a representative individual in the group exposed to the greatest risk. SSM's regulations also imply that the assessment time for a repository of this type is one million years after closure. The licence applied for is one in a stepwise series of permits, each requiring a safety report. The next step concerns a permit to start excavation of the repository and requires a preliminary safety assessment report (PSAR) covering both operational and post-closure safety. Later steps include permission to commence trial operation, to commence regular operation and to close the final repository. (authors)

Transition to Office-based Obstetric and Gynecologic Procedures: Safety, Technical, and Financial Considerations.

Science.gov (United States)

Peacock, Lisa M; Thomassee, May E; Williams, Valerie L; Young, Amy E

2015-06-01

Office-based surgery is increasingly desired by patients and providers due to ease of access, overall efficiency, reimbursement, and satisfaction. The adoption of office-based surgery requires careful consideration of safety, efficacy, cost, and feasibility within a providers practice. This article reviews the currently available data regarding patient and provider satisfaction as well as practical considerations of staffing, equipment, and supplies. To aid the practitioner, issues of office-based anesthesia and safety with references to currently available national guidelines and protocols are provided. Included is a brief review of billing, coding, and reimbursement. Technical procedural aspects with information and recommendations are summarized.

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of real component geometries (phase 2). Final report; Erhoehung der Komponentensicherheit durch verbesserte Verfahren zur Eigenspannungsanalyse. Teilvorhaben. Beruecksichtigung realer Komponentengeometrien (Phase 2). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Scholtes, B.; Nau, Andreas

2015-06-01

Residual stresses can be result in both detrimental as well as beneficial consequences on the component's strength and lifetime. A most detailed knowledge of the residual stress state is a pre-requisite for the assessment of the component's performance. The mechanical methods for residual stress measurements are classified in non-destructive, destructive and semi-destructive methods. The two commonly used (semi-destructive) mechanical methods are the hole drilling and the ring core method. In the context of reactor safety research of the Federal Ministry of Economic Affairs and Energy (BMWi), two fundamental and interacting weak points of the hole drilling as well as ofthe ring core method are investigated. On the one hand, there are effects concerning geometricalboundary conditions of the components and on the other hand, there are influences of plasticity due to notch effects. Both aspects affect the released strain field, when the material is removed and finally, the calculated residual stresses. The first issue mentioned above is under the responsibility of Institute of Materials Engineering - Metallic Materials (KasselUniversity) and the last one will be investigated by University of Stuttgart-Otto-Graf-Institut - materials testing institute. As a consequence of a successful project, the present knowledgebase will be considerably improved and will be available for various engineering fields. Especially,the quantitative consideration of real residual stress states for optimized component designs will be feasible and finally the consequences of residual stresses on the component's safety, which are used in nuclear facilities, can be evaluated. The findings of the application-oriented research period (phase 2) at Kassel University are documented in this report.

9 CFR 202.115 - Rule 15: Submission for final consideration.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Rule 15: Submission for final consideration. 202.115 Section 202.115 Animals and Animal Products GRAIN INSPECTION, PACKERS AND STOCKYARDS... the judicial officer for decision. (d) Oral argument. There shall be no right to oral argument other...

Hawaii State Plan for Occupational Safety and Health. Final rule.

Science.gov (United States)

2012-09-21

This document announces the Occupational Safety and Health Administration's (OSHA) decision to modify the Hawaii State Plan's ``final approval'' determination under Section 18(e) of the Occupational Safety and Health Act (the Act) and to transition to ``initial approval'' status. OSHA is reinstating concurrent federal enforcement authority over occupational safety and health issues in the private sector, which have been solely covered by the Hawaii State Plan since 1984.

Safety and security considerations for the transport of spent teletherapy units

International Nuclear Information System (INIS)

Mallaupoma, Mario; Paez, Jose; Huatay, Luis; Cruz, Walter

2008-01-01

Among the applications of nuclear technology, a practice widely used and generates many benefits to society are teletherapy applications. Many of the teletherapy units used contain a source of cobalt-60 and after their useful life they have to be dismantled and transported to a safe place. In this case were transported two units with an activity of more than 75 TBq . This paper presents safety and security considerations for the transport of the teletherapy units according to the recommendations of actual state of art. It is described all facets of safe transport by means of a set of technical and administrative safety requirements and controls, including the actions required by the consignor and carrier. The main emphasis was put on the stages of transport operations that give rise to exposure to radiation like packing, preparation, loading, handling, storage in transit and movement of packages of radioactive material. On the other side some security actions were considered in order to prevent theft, sabotage or other malicious acts during the transport of the packages. As a conclusion it must be mentioned that both safety and security considerations are very important aspects that must be taking in account for the transport of high activity radioactive material. (author)

Defining safety goals. 2. Basic Consideration on Defining Safety Goals

International Nuclear Information System (INIS)

Hakata, T.

2001-01-01

The purpose of this study is to develop basic safety goals that are rational and consistent for all nuclear facilities, including nuclear power plants and fuel cycle facilities. Basic safety goals (risk limits) by an index of radiation dose are discussed, which are based on health effects of detriment and fatality and risk levels presumably accepted by society. The contents of this paper are the personal opinions of the author. The desirable structure of safety goals is assumed to be 'basic safety goals plus specific safety goals (or supplemental safety goals) for each sort of facility, which reflects their characteristics'. The requisites of the basic safety goals must include (a) rational bases (scientific and social), (b) comprehensiveness (common to all sorts of nuclear facilities covering from normal to accidental conditions), and (c) applicability. To meet the requirements, the basic safety goals might have to be a risk profile expression by an index of radiation dose. The societal rationality is consideration of absolute risk levels (10 -6 or 10 -7 /yr) and/or relative risk factors (such as 0.1% of U.S. safety goals) that the general public accepts as tolerable. The following quantitative objectives are adopted in this study for protection of average individuals in the vicinity of a nuclear facility: 1. The additive annual radiation dose during normal operation must be -4 /yr (health detriment), 2x10 -6 /yr (latent cancer and severe hereditary effects), and 10 -7 /yr (acute fatality) from the statistics in Japan. The radiation effects on human beings are determined by recommendations of UNSCEAR (Ref. 1) and ICRP. The health effects considered are non-severe stochastic health detriment, i.e., detectable opacities of lens of eye (threshold 5 0.5 to 2 Sv), depression of hematopoiesis of bone marrow (0.5 Sv), and depression of reproductive capability (temporary sterility of testes ) (0.15 Sv). The LD 50/60 of acute fatality is ∼4 Sv, and fatalities by latent

Safety considerations of PWR's

International Nuclear Information System (INIS)

Arnold, W.H. Jr.

1977-01-01

The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

Improving the safety of LWR power plants. Final report

International Nuclear Information System (INIS)

1980-04-01

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

Safety considerations in the design of an electron beam melter for a fissile metal

International Nuclear Information System (INIS)

Jenkins, T.W.

1995-01-01

The purpose of the EBM in the melting of highly enriched uranium metal is principally to recycle machining swarf but it may also be used for the manufacture of alloys and the recycling of alloy swarf. The current recycling of swarf involves the cleaning and compaction of swarf to pellets, these are then melted in a VIF with losses of > 20% of the metal as dross lift in the crucible. This has to be converted to oxide and then chemically recycled, a very expensive process. Use of the EBM would limit losses to less than 3% and make significant savings and also solve some safety problems. The EBM will be attached to a glove box and fume cupboard and the hazards addressed by the design include: (1) Criticality, (2) Radiation, (3) Release of Radioactive Contamination, (4) Explosion, (5) Fire, (6) Mechanical Handling, (7) Electrical, (8) Other Safety Considerations. These are addressed with details of the considerations including interlocks required to mitigate the hazards. The requirements of Safety documentation and Hazard and Operability studies are outlined together with Quality Assurance demands and training requirements. A number of safety considerations are based on previous accident scenarios in which Dangerous Occurrences took place involving equipment faults and operator errors on high vacuum equipment associated with glove boxes in a radioactive controlled area

Application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal. Final report; Anwendung und Weiterentwicklung von Modellen fuer Endlagersicherheitsanalysen auf die Freigabe radioaktiver Stoffe zur Deponierung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Artmann, Andreas; Larue, Juergen; Seher, Holger; Weiss, Dietmar

2014-08-15

The project of application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal is aimed to study the long-term safety using repository-specific simulation programs with respect to radiation exposure for different scenarios. It was supposed to investigate whether the 10 micro Sv criterion can be guaranteed under consideration of human intrusion scenarios. The report covers the following issues: selection and identification of models and codes and the definition of boundary conditions; applicability of conventional repository models for long-term safety analyses; modeling results for the pollutant release and transport and calculation of radiation exposure; determination of the radiation exposure.

Ferrocyanide safety project ferrocyanide aging studies. Final report

International Nuclear Information System (INIS)

Lilga, M.A.; Hallen, R.T.; Alderson, E.V.

1996-06-01

This final report gives the results of the work conducted by Pacific Northwest National Laboratory (PNNL) from FY 1992 to FY 1996 on the Ferrocyanide Aging Studies, part of the Ferrocyanide Safety Project. The Ferrocyanide Safety Project was initiated as a result of concern raised about the safe storage of ferrocyanide waste intermixed with oxidants, such as nitrate and nitrite salts, in Hanford Site single-shell tanks (SSTs). In the laboratory, such mixtures can be made to undergo uncontrolled or explosive reactions by heating dry reagents to over 200 degrees C. In 1987, an Environmental Impact Statement (EIS), published by the U.S. Department of Energy (DOE), Final Environmental Impact Statement, Disposal of Hanford Defense High-Level Transuranic and Tank Waste, Hanford Site, Richland, Washington, included an environmental impact analysis of potential explosions involving ferrocyanide-nitrate mixtures. The EIS postulated that an explosion could occur during mechanical retrieval of saltcake or sludge from a ferrocyanide waste tank, and concluded that this worst-case accident could create enough energy to release radioactive material to the atmosphere through ventilation openings, exposing persons offsite to a short-term radiation dose of approximately 200 mrem. Later, in a separate study (1990), the General Accounting Office postulated a worst-case accident of one to two orders of magnitude greater than that postulated in the DOE EIS. The uncertainties regarding the safety envelope of the Hanford Site ferrocyanide waste tanks led to the declaration of the Ferrocyanide Unreviewed Safety Question (USQ) in October 1990

Study of fundamental safety-related aspects in connection with the decommissioning of nuclear installations. Pt. 2. Safety considerations and emissions

International Nuclear Information System (INIS)

John, T.; Thierfeldt, S.

1993-01-01

The procedures used so far for the examination of selected decommissioning projects in expert opinions on safety, in particular of nuclear power plants, were screened, with special emphasis on the examination of safety considerations, i.e. analysis of possible accidents. Generic examinations on safety in connection with the decommissioning of nuclear installations were used to assess safety considerations. Different approaches were taken with regard to the selection of analysed accidents and determination of parameters defining activity release and assumptions in safety opinions. Therefore it seems to be appropriate to establish a scenario to be used for nuclear power plant accident analyses, which covers the range of radiologically relevant accidents during decommissioning activities. Although it might be controversially discussed, because of specific plant designs (test and prototype reactors as well as first power reactors), to establish such a radiologically covering accident scenario for older nuclear power plants, it seems to be no problem for modern light water reactors. The radiologically most relevant possible accident in a decommissioned nuclear power plant is fire in the plant. Parameter values and assumptions are suggested which determine the source term in the event of a fire in the plant. Inspite of a conservative determination of parameter values and assumptions, an environmental dose commitment of less than 50 mSv is to be expected for the resulting source term. (orig.) [de

Considerations for the nonclinical safety evaluation of antibody drug conjugates for oncology.

Science.gov (United States)

Roberts, Stanley A; Andrews, Paul A; Blanset, Diann; Flagella, Kelly M; Gorovits, Boris; Lynch, Carmel M; Martin, Pauline L; Kramer-Stickland, Kimberly; Thibault, Stephane; Warner, Garvin

2013-12-01

Antibody drug conjugates (ADCs) include monoclonal antibodies that are linked to cytotoxic small molecules. A number of these agents are currently being developed as anti-cancer agents designed to improve the therapeutic index of the cytotoxin (i.e., cytotoxic small molecule or cytotoxic agent) by specifically delivering it to tumor cells. This paper presents primary considerations for the nonclinical safety evaluation of ADCs and includes strategies for the evaluation of the entire ADC or the various individual components (i.e., antibody, linker or the cytotoxin). Considerations are presented on how to design a nonclinical safety assessment program to identify the on- and off-target toxicities to enable first-in-human (FIH) studies. Specific discussions are also included that provide details as to the need and how to conduct the studies for evaluating ADCs in genetic toxicology, tissue cross-reactivity, safety pharmacology, carcinogenicity, developmental and reproductive toxicology, biotransformation, toxicokinetic monitoring, bioanalytical assays, immunogenicity testing, test article stability and the selection of the FIH dose. Given the complexity of these molecules and our evolving understanding of their properties, there is no single all-encompassing nonclinical strategy. Instead, each ADC should be evaluated on a case-by-case scientifically-based approach that is consistent with ICH and animal research guidelines. Copyright © 2013 Elsevier Inc. All rights reserved.

Safety Parameter Considerations of Anodal Transcranial Direct Current Stimulation in Rats

Science.gov (United States)

2017-10-01

Richardson, J.D., Baker, J.M., Rorden, C., 2011. Transcranial direct current stimulation improves naming reaction time in fluent aphasia: a...AFRL-RH-WP-TR-2017-0069 Safety parameter considerations of anodal transcranial Direct Current Stimulation in rats R. Andy McKinley...response, including the time for reviewing instructions, searching existing data sources, searching existing data sources, gathering and maintaining the

GENERAL CONSIDERATIONS ON REGULATIONS AND SAFETY REQUIREMENTS FOR QUADRICYCLES

Directory of Open Access Journals (Sweden)

Ana Pavlovic

2015-12-01

Full Text Available In recent years, a new class of compact vehicles has been emerging and wide-spreading all around Europe: the quadricycle. These four-wheeled motor vehicles, originally derived from motorcycles, are a small and fuel-efficient mean of transportation used in rural or urban areas as an alternative to motorbikes or city cars. In some countries, they are also endorsed by local authorities and institutions which support small and environmentally-friendly vehicles. In this paper, several general considerations on quadricycles will be provided including the vehicle classification, evolution of regulations (as homologation, driver licence, emissions, etc, technical characteristics, safety requirements, most relevant investigations, and other additional useful information (e.g. references, links. It represents an important and actual topic of investigation for designers and manufacturers considering that the new EU regulation on the approval and market surveillance of quadricycles will soon enter in force providing conclusive requirements for functional safety environmental protection of these promising vehicles.

Considerations concerning the reliability of reactor safety equipment

International Nuclear Information System (INIS)

Furet, J.; Guyot, Ch.

1967-01-01

A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [fr

76 FR 53051 - Safety Zone; ISAF Nations Cup Grand Final Fireworks Display, Sheboygan, WI

Science.gov (United States)

2011-08-25

... display. DATES: This rule is effective from 7:45 until 8:45 p.m. on September 13, 2011. ADDRESSES...-AA00 Safety Zone; ISAF Nations Cup Grand Final Fireworks Display, Sheboygan, WI AGENCY: Coast Guard, DHS. ACTION: Temporary final rule. SUMMARY: The Coast Guard is establishing a temporary safety zone on...

Project Guarantee 1985. Final repository for low- and intermediate level radioactive wastes: Safety report

International Nuclear Information System (INIS)

Anon.

1985-01-01

Storage of radioactive waste must delay the return of radionuclides to the biosphere for a long period of time and must maintain the release rates at a sufficiently low level for all time. This is achieved with the aid of a series of safety barriers which consist, on the one hand, of technical barriers in the repository and, on the other hand , of natural geological barriers as they occur at the repository location. In order to assess the efficiency of the barriers, the working methods of the technical barriers and the host rock must be understood. This understanding is transferred into quantitative models in order to calculate the safety of the repository. The individual barriers and the methods used to modelling their functions were described in volume NGB 85-07 of the Project Guarantee 1985 report series and the data necessary for modelling were given. The models and data are used in the safety analysis, the results of which are contained in the present report. Safety considerations show that models are available in Switzerland which allow, in principle, an assessment of the long-term behaviour of a repository for low- and intermediate-level waste. The evaluation of earlier studies and experimental work, suitable laboratory measurements and results from field research enable compilation of a representative data-set so that the requirements for quantitative statements on safety of final disposal are met from this side also. The safety calculations show that the radiation doses calculated for a base case scenario with realistic/conservative parameter values are negligibly low. Also, radiation doses which are clearly under the protection standard of 10 mrem per year result for conservative values and the cumulation of several conservative assumptions. Even assuming exposure of the repository by erosion, a radiotoxicity of the soil formed results which is under natural values

SKB 91. Final disposal of spent nuclear fuel. Importance of the bedrock for safety

International Nuclear Information System (INIS)

1992-05-01

The safety of a deep repository for spent nuclear fuel has been assessed in this report. The spent fuel is assumed to be encapsulated in a copper canister and deposited at a depth of 600 m in the bedrock. The primary purpose has been to shed light on the importance of the geological features of the site for the safety of a final repository. The assessment shows that the encapsulated fuel will, in all likelihood, be kept isolated from the groundwater for millions of years. This is considerably longer than the more than 100 000 years that are required in order for the toxicity of the waste to have declined to a level equivalent to that of rich uranium ores. However, in order to be able to study the role of the rock as a barrier to the dispersal of radioactive materials, calculations have been carried out under the assumption that waste canisters leak. The results show that the safety of a carefully designed repository is only affected to a small extent by the ability of the rock to retain the escaping radionuclides. The primary role of the rock is to provide stable mechanical and chemical conditions in the repository over a long period of time so that the function of the engineered barriers is not jeopardized. (187 refs.) (au)

Consideration of timescales in post-closure safety of geological disposal of radioactive waste

International Nuclear Information System (INIS)

2006-11-01

by a number of general considerations, which are described first. Three broad areas in the regulation and practice of repository planning and implementation affected by timescales issues are then discussed: - repository siting and design and the levels of protection required in regulation; - the planning of pre- and post-closure actions; and - developing and presenting a safety case. Finally, a synthesis of findings is made, including a review of the statements made in the 2004 'lessons learnt' report in light of the discussions contained in the present report. Many of the issues treated in the course of the project are subject to various interpretations, and remain under discussion in national programmes, as well as internationally. Therefore, the findings in this report should not be viewed as conclusive, but rather as a contribution in moving ahead the debate and understanding the similarities and differences among approaches in national programmes. (authors)

Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

International Nuclear Information System (INIS)

Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

1990-01-01

FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

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

Science.gov (United States)

Haslberger, Alexander G

2006-05-03

information. An integrated assessment might help to focus and save capacities in highly technical areas such as molecular characterization or profiling, which are often necessary for both assessments. In the area of establishing international standards for traded foods, such as for the newly created Standards in Trade and Development Facility (STDF), an integrated assessment might help in the consideration of important environmental aspects involved in health and food safety. Furthermore, an established integrated view on GMOs may create greater consumer confidence in the technology.

Jefferson Lab IEC 61508/61511 Safety PLC Based Safety System

International Nuclear Information System (INIS)

Mahoney, Kelly; Robertson, Henry

2009-01-01

This paper describes the design of the new 12 GeV Upgrade Personnel Safety System (PSS) at the Thomas Jefferson National Accelerator Facility (TJNAF). The new PSS design is based on the implementation of systems designed to meet international standards IEC61508 and IEC 61511 for programmable safety systems. In order to meet the IEC standards, TJNAF engineers evaluated several SIL 3 Safety PLCs before deciding on an optimal architecture. In addition to hardware considerations, software quality standards and practices must also be considered. Finally, we will discuss R and D that may lead to both high safety reliability and high machine availability that may be applicable to future accelerators such as the ILC.

Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

International Nuclear Information System (INIS)

Reyes, Susana; Anklam, Tom; Meier, Wayne; Campbell, Patrick; Babineau, Dave; Becnel, James; Taylor, Craig; Coons, Jim

2016-01-01

Highlights: • The safety characteristics and at risk inventories in an IFE facility are discussed. • The primary nuclear hazard is the potential exposure of workers and/or the public to tritium and/or neutronically activated products. • Recent technology developments in tritium processing are key for minimization of inventories. • Initial safety studies indicate that hazards associated to the use of liquid lithium can be appropriately managed. • Simulation of worst-case scenarios indicate that the accident consequences are limited and below the limit for public evacuation. - Abstract: Over the past five years, the fusion energy group at Lawrence Livermore National Laboratory (LLNL) has made significant progress in the area of safety and tritium research for Inertial Fusion Energy (IFE). Focus has been driven towards the minimization of inventories, accident safety, development of safety guidelines and licensing considerations. Recent technology developments in tritium processing and target fill have had a major impact on reduction of tritium inventories in the facility. A safety advantage of inertial fusion energy using indirect-drive targets is that the structural materials surrounding the fusion reactions can be protected from target emissions by a low-pressure chamber fill gas, therefore eliminating plasma-material erosion as a source of activated dust production. An important inherent safety advantage of IFE when compared to other magnetic fusion energy (MFE) concepts that have been proposed to-date (including ITER), is that loss of plasma control events with the potential to damage the first wall, such as disruptions, are non-conceivable, therefore eliminating a number of potential accident initiators and radioactive in-vessel source term generation. In this paper, we present an overview of the safety assessments performed to-date, comparing results to the US DOE Fusion Safety Standards guidelines and the recent lessons-learnt from ITER safety and

Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

Energy Technology Data Exchange (ETDEWEB)

Reyes, Susana, E-mail: reyes20@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Anklam, Tom; Meier, Wayne; Campbell, Patrick [Lawrence Livermore National Laboratory, Livermore, CA (United States); Babineau, Dave; Becnel, James [Savannah River National Laboratory, Aiken, SC (United States); Taylor, Craig; Coons, Jim [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-11-01

Highlights: • The safety characteristics and at risk inventories in an IFE facility are discussed. • The primary nuclear hazard is the potential exposure of workers and/or the public to tritium and/or neutronically activated products. • Recent technology developments in tritium processing are key for minimization of inventories. • Initial safety studies indicate that hazards associated to the use of liquid lithium can be appropriately managed. • Simulation of worst-case scenarios indicate that the accident consequences are limited and below the limit for public evacuation. - Abstract: Over the past five years, the fusion energy group at Lawrence Livermore National Laboratory (LLNL) has made significant progress in the area of safety and tritium research for Inertial Fusion Energy (IFE). Focus has been driven towards the minimization of inventories, accident safety, development of safety guidelines and licensing considerations. Recent technology developments in tritium processing and target fill have had a major impact on reduction of tritium inventories in the facility. A safety advantage of inertial fusion energy using indirect-drive targets is that the structural materials surrounding the fusion reactions can be protected from target emissions by a low-pressure chamber fill gas, therefore eliminating plasma-material erosion as a source of activated dust production. An important inherent safety advantage of IFE when compared to other magnetic fusion energy (MFE) concepts that have been proposed to-date (including ITER), is that loss of plasma control events with the potential to damage the first wall, such as disruptions, are non-conceivable, therefore eliminating a number of potential accident initiators and radioactive in-vessel source term generation. In this paper, we present an overview of the safety assessments performed to-date, comparing results to the US DOE Fusion Safety Standards guidelines and the recent lessons-learnt from ITER safety and

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.

Nuclear safety and radiation protection consideration in the design of research and development facility

International Nuclear Information System (INIS)

Akbar, M.R.

2010-01-01

Nuclear safety is a critically important aspect that must be considered in the design of a nuclear facility in order to ensure the protection of the workers, public and environment. This paper looks at the methodology, approach and incorporation of this aspect, specifically into the design of a research and development facility. The Health, Safety and Environmental Basis of Design is an initial analysis of nuclear safety and radiation protection considerations that is performed during the conceptual design phase and sets the baseline for what the design of the facility must conform to. It consists of general nuclear safety design principles, such as defence in depth and optimisation considerations, and a hazard management strategy. Following the Health, Safety and Environmental Basis of Design, a Preliminary Safety Assessment Report is generated during the basic design phase in conjunction with various analyses in order to assess the impact of hazards on the workers and members of the public. This assessment follows a hazard graded approach where the depth of the analysis will be determined by the impact of the worst case accident scenario in the facility. The assessment also includes a waste management strategy which is an essential aspect to be considered in the design in order to minimize the generation of waste. The safety assessment also demonstrates compliance to dose limits and risk criteria for the workers and members of the public set by the regulatory body and supported by a legal framework. Measures are taken to keep risk as low as reasonably achievable and prevent transgression of the risk and dose limits. However, a balance needs to be maintained between 5 reducing these doses further and the cost of such a reduction, which is known as optimization. It is therefore imperative to have nuclear safety specialists analyse the design in order to protect the worker and member of the public from unwarranted exposure to nuclear radiation. (author)

Safety technical considerations on the 2012 periodic safety verification of the Beznau nuclear power plant

International Nuclear Information System (INIS)

2016-12-01

According to nuclear legislation, the owner of an operational license for a nuclear power plant has to provide a periodic safety verification (PSU) every 10 years. The 'North Eastern Power Plants' company (NOK), today AXPO Power AG already performed such a PSU for the Beznau-2 nuclear reactor block (KKB2) in 2002. The Beznau-1 nuclear reactor block (KKB1) received its definitive operational license in October 1970, after test operation during 7 months. After the license for test operation received on July 16 th , 1971, the operational license of KKB2 was renewed several times, each time for a certain period of validity. In 1991, NOK requested a definitive operational license for KKB2, but in 1994 the Swiss Federal Council lengthened the license for only 10 years. Moreover, it laid down that NOK has to periodically report on the safety of the facility. With its letter of August 23 rd , 1998, the Federal Office of Energy defined the documents to be produced for the PSU. The extent of the PSU was defined in such a way that many documents concern the whole power plant, i.e. both nuclear reactor blocks. On December 3 rd , 2004, the Swiss Federal Council granted KKB2 an operational license of limited validity. The present report reviews the 2012 PSU, which covers the time interval from January 1 st , 2002, to December 31 st , 2011, from the point of view of safety. It contains documents for the evaluation of both reactor blocks at KKB. The Beznau interim storage pool was also taken into consideration; it is situated on the KKB site, but, according to a decision of the Swiss Federal Council of May 23 rd , 1991, it has an independent operational license. The evaluation of ageing surveillance takes the whole operational period of the facility into account, i.e. the ageing mechanisms acting as from the beginning of the operation. Moreover, important developments that occurred after the surveillance time interval have been taken into account, especially the status

A consideration of lithium cell safety

Science.gov (United States)

Tobishima, Shin-ichi; Yamaki, Jun-ichi

The safety characteristics of commercial lithium ion cells are examined in relation to their use as batteries for cellular phones. This report describes a theoretical approach to an understanding of cell safety, example results of safety tests that we performed on lithium ion cells, and also presents our views regarding cell safety.

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Complementary considerations 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

Complementary Considerations sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of enhancing confidence in the outcomes of the safety assessment for a spent nuclear fuel repository to be constructed at Olkiluoto, Finland. The main emphasis in this report is on the evidence and understanding that can be gained from observations at the site, including its regional geological environment, and from natural and anthropogenic analogues for the repository, its components and the processes that affect safety. In particular, the report addresses diverse and less quantifiable types of evidence and arguments that are enclosed to enhance confidence in the outcome of the safety assessment. These complementary considerations have been described as evaluations, evidence and qualitative supporting arguments that lie outside the scope of the other reports of the quantitative safety assessment. The experience with natural analogues for the long-term durability of the materials involved and the extent of processes provides high confidence in our understanding of the disposal system and its evolution. For each engineered barrier and key process, there is increasing analogue evidence to support the conceptual models and parameters. Regarding the suitability of the Olkiluoto site to host a spent fuel repository, a number of factors have been identified that indicate the suitability of crystalline host rock in general, and that of the Olkiluoto site in particular. The report also provides radiation background information for the use of complementary indicators, which aid in putting the results of the safety analysis presented in Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere Assessment in a broader perspective to show that the radiation originating from a spent nuclear fuel repository remains in most cases much below natural background radiation or that caused by non-nuclear industries. (orig.)

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Complementary considerations 2012

International Nuclear Information System (INIS)

2012-12-01

Complementary Considerations sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of enhancing confidence in the outcomes of the safety assessment for a spent nuclear fuel repository to be constructed at Olkiluoto, Finland. The main emphasis in this report is on the evidence and understanding that can be gained from observations at the site, including its regional geological environment, and from natural and anthropogenic analogues for the repository, its components and the processes that affect safety. In particular, the report addresses diverse and less quantifiable types of evidence and arguments that are enclosed to enhance confidence in the outcome of the safety assessment. These complementary considerations have been described as evaluations, evidence and qualitative supporting arguments that lie outside the scope of the other reports of the quantitative safety assessment. The experience with natural analogues for the long-term durability of the materials involved and the extent of processes provides high confidence in our understanding of the disposal system and its evolution. For each engineered barrier and key process, there is increasing analogue evidence to support the conceptual models and parameters. Regarding the suitability of the Olkiluoto site to host a spent fuel repository, a number of factors have been identified that indicate the suitability of crystalline host rock in general, and that of the Olkiluoto site in particular. The report also provides radiation background information for the use of complementary indicators, which aid in putting the results of the safety analysis presented in Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere Assessment in a broader perspective to show that the radiation originating from a spent nuclear fuel repository remains in most cases much below natural background radiation or that caused by non-nuclear industries. (orig.)

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Complementary considerations 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

Complementary Considerations sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of enhancing confidence in the outcomes of the safety assessment for a spent nuclear fuel repository to be constructed at Olkiluoto, Finland. The main emphasis in this report is on the evidence and understanding that can be gained from observations at the site, including its regional geological environment, and from natural and anthropogenic analogues for the repository, its components and the processes that affect safety. In particular, the report addresses diverse and less quantifiable types of evidence and arguments that are enclosed to enhance confidence in the outcome of the safety assessment. These complementary considerations have been described as evaluations, evidence and qualitative supporting arguments that lie outside the scope of the other reports of the quantitative safety assessment. The experience with natural analogues for the long-term durability of the materials involved and the extent of processes provides high confidence in our understanding of the disposal system and its evolution. For each engineered barrier and key process, there is increasing analogue evidence to support the conceptual models and parameters. Regarding the suitability of the Olkiluoto site to host a spent fuel repository, a number of factors have been identified that indicate the suitability of crystalline host rock in general, and that of the Olkiluoto site in particular. The report also provides radiation background information for the use of complementary indicators, which aid in putting the results of the safety analysis presented in Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere Assessment in a broader perspective to show that the radiation originating from a spent nuclear fuel repository remains in most cases much below natural background radiation or that caused by non-nuclear industries. (orig.)

Technical considerations for the development of an engineering safety features control system with PLC

International Nuclear Information System (INIS)

Lee, C. K.; Kim, C. H.; Han, J. B.; Kim, H.; Lee, S. S.

2002-01-01

Technical considerations are summarized for the development of an ESFCS(Engineered Safety Features Control System) with PLC (Programmable Logic Controller). The ESFCS is required for the mitigation of plant accident conditions and therefore developed in conformance with the design requirements applied to the safety critical system. The design of ESFCS primarily considered its safety, and the system has an architecture that will be able to minimize spurious actuation. The PLC based functional distribution and redundant design features are adopted, and the fieldbus is applied in the communication of information and control signals between PLC processors. It is expected that the ESFCS will have several advanced design features compared with the conventional systems supplied by foreign vendors

An assessment of traffic safety culture related to engagement in efforts to improve traffic safety : final report.

Science.gov (United States)

2016-12-01

This final report summarizes the methods, results, conclusions, and recommendations derived from a survey conducted to understand values, beliefs, and attitudes regarding engagement in behaviors that impact the traffic safety of others. Results of th...

Prioritization of tasks in the draft LWR safety technology program plan. Final report

International Nuclear Information System (INIS)

Lim, E.Y.; Miller, W.J.; Parkinson, W.J.; Ritzman, R.L.; vonHerrmann, J.L.; Wood, P.J.

1980-05-01

The purpose of this report is to describe both the approach taken and the results produced in the SAI effort to prioritize the tasks in the Sandia draft LWR Safety Technology Program Plan. This work used the description of important safety issues developed in the Reactor Safety Study (2) to quantify the effect of safety improvements resulting from a research and development program on the risk from nuclear power plants. Costs of implementation of these safety improvements were also estimated to allow a presentation of the final results in a value (i.e., risk reduction) vs. impact (i.e., implementation costs) matrix

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of the elastic-plastic material properties (Phase 2). Final report

International Nuclear Information System (INIS)

Mirbach, David von

2015-01-01

Residual stresses in mechanical components can result in both detrimental but also beneficial effects on the strength and lifetime of the components. The most detailed knowledge of the residual stress state is of advantage or a pre-requisite for the assessment of the component performance. Two commonly used methods for determination of residual stresses are the hole drilling method and the ring core method which can be regarded to the mechanical methods. In the context of reactor safety research of the German Federal Ministry of Economic and Energy (BMWi) two fundamental and interacting weak points of the hole drilling method as well as of the ring core method, respectively, in order to determine residual stresses are going to be investigated. As a consequence reliability of the methods will be improved in this joint research project. On the one hand there are effects of geometrical boundary conditions of the components and on the other hand there is the influence of plasticity due to notch effects both affecting the released strain field after removing material and after all the calculated residual stresses. The first issue mentioned above is under the responsibility of the Institute of Materials Engineering (Kassel University) and the last one is investigated by materials testing institute university Stuttgart. As a consequence of a successful project the knowledge base will be considerably improved resulting in benefits for various engineering fields. Especially the quantitative consideration of real residual stress states for optimized component designs will be possible and after all the consequences of residual stresses on safety of components which are used in nuclear facilities can be evaluated. In this second experimental research chapter (phase 2) the findings of the first numerical and theoretical research chapter (phase 1) where proofed. The developed differential calculation method with the method of adaptive calibration functions were compared with the

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of the elastic-plastic material properties (phase 1). Final report

International Nuclear Information System (INIS)

Mirbach, David von

2014-01-01

Residual stresses in mechanical components can result in both detrimental but also beneficial effects on the strength and lifetime of the components. The most detailed knowledge of the residual stress state is of advantage or a pre-requisite for the assessment of the component performance. The mechanical methods for residual stress measurement are divided into the groups of non-destructive and destructive methods. Two commonly used mechanical methods for determination of residual stresses are the hole drilling method and the ring core method which can be regarded as semi-destructive methods. In the context of reactor safety research of the German Federal Ministry of Economic and Technology (BMWi) two fundamental and interacting weak points of the hole drilling method as well as of the ring core method, respectively, in order to determine residual stresses are going to be investigated. As a consequence reliability of the methods will be improved in this joint research project. On the one hand there are effects of geometrical boundary conditions of the components and on the other hand there is the influence of plasticity due to notch effects both affecting the released strain field after removing material and after all the calculated residual stresses. The first issue mentioned above is under the responsibility of the Institute of Materials Engineering (Kassel University) and the last one is investigated by Universitaet of Stuttgart-Otto-Graf-Institut - materials testing institute. As a consequence of a successful project the knowledge base will be considerably improved resulting in benefits for various engineering fields. Especially the quantitative consideration of real residual stress states for optimized component designs will be possible and after all the consequences of residual stresses on safety of components which are used in nuclear facilities can be evaluated. The state of art was reground in the first research chapter and the analysed strain gauges where

Ab initio chemical safety assessment: A workflow based on exposure considerations and non-animal methods

OpenAIRE

Berggren, Elisabet; White, Andrew; Ouedraogo, Gladys; Paini, Alicia; Richarz, Andrea-Nicole; Bois, Frederic Y.; Exner, Thomas; Leite, Sofia; Grunsven, Leo A. van; Worth, Andrew; Mahony, Catherine

2017-01-01

Highlights • A workflow for an exposure driven chemical safety assessment to avoid animal testing. • Hypothesis based on existing data, in silico modelling and biokinetic considerations. • A tool to inform targeted and toxicologically relevant in vitro testing.

Safety research colloquium 2013-2014. Vol. 10

International Nuclear Information System (INIS)

Pieper, Ralf

2015-01-01

Volume 10 of the safety research colloquium 2013-2014 covers the following issues: Design, ergonomics and safety in product development; Germany is searching a final repository site: concepts and status of the final disposal of nuclear waste; collaborating robots - status of research, standardization and validation; psychological workloads - empirical indications; psychological workloads - actual challenges; expert security by occupational health management - challenges to operational practice; expert security by occupational health management - example of a demographic program in the practical realization; challenges in employment legislation - reduction of the key staff; consideration of human factors in hazard assessment a a challenge for every safety engineer, innovative technologies for work equipment and working systems in the context of ambient intelligence and industry 4.0; challenges of functional safety in the automotive sector; nanotechnology - an example for successful technology assessment.

An overview of safety and environmental considerations in the selection of materials for fusion facilities

International Nuclear Information System (INIS)

Petti, D.A.; Piet, S.J.; Seki, Y.

1996-01-01

Safety and environmental considerations can play a large role in the selection of fusion materials. In this paper, we review the attributes of different structural, plasma facing, and breeding materials from a safety perspective and discuss some generic waste management issues as they relate to fusion materials in general. Specific safety concerns exist for each material that must be dealt with in fusion facility design. Low activation materials offer inherent safety benefits compared with conventional materials, but more work is needed before these materials have the requisite certified databases. In the interim, the international thermonuclear experimental reactor (ITER) has selected more conventional materials and is showing that the safety concerns with these materials can be addressed by proper attention to design. In the area of waste management disposal criteria differ by country. However, the criteria are all very strict making disposal of fusion components difficult. As a result, recycling has gained increasing attention. (orig.)

Waste management safety

International Nuclear Information System (INIS)

Boehm, H.

1983-01-01

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

Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 92--94). Final report

International Nuclear Information System (INIS)

Ho, S.K.; Fowler, T.K.; Holdren, J.P.

1994-01-01

This is the Final Report for a three-year (FY 92--94) study of the Environmental, Safety, and Economic (ESE) aspects of fusion energy systems, emphasizing development of computerized approaches suitable for incorporation as modules in fusion system design codes. First, as is reported in Section 2, the authors now have operating a simplified but complete environment and safety evaluation code, BESAFE. The first tests of BESAFE as a module of the SUPERCODE, a design optimization systems code at LLNL, are reported in Section 3. Secondly, as reported in Section 4, the authors have maintained a strong effort in developing fast calculational schemes for activation inventory evaluation. In addition to these major accomplishments, considerable progress has been made on research on specific topics as follows. A tritium modeling code TRIDYN was developed in collaboration with the TSTA group at LANL and the Fusion Nuclear Technology group at UCLA. A simplified algorithm has been derived to calculate the transient temperature profiles in the blanket during accidents. The scheme solves iteratively a system of non-linear ordinary differential equations describing about 10 regions of the blanket by preserving energy balance. The authors have studied the physics and engineering aspects of divertor modeling for safety applications. Several modifications in the automation and characterization of environmental and safety indices have been made. They have applied this work to the environmental and safety comparisons of stainless steel with alternative structural materials for fusion reactors. A methodology in decision analysis utilizing influence and decision diagrams has been developed to model fusion reactor design problems. Most of the work during this funding period has been reported in 26 publications including theses, journal publications, conference papers, and technical reports, as listed in Section 11

Using of BEPU methodology in a final safety analysis report

International Nuclear Information System (INIS)

Menzel, Francine; Sabundjian, Gaiane; D'auria, Francesco; Madeira, Alzira A.

2015-01-01

The Nuclear Reactor Safety (NRS) has been established since the discovery of nuclear fission, and the occurrence of accidents in Nuclear Power Plants worldwide has contributed for its improvement. The Final Safety Analysis Report (FSAR) must contain complete information concerning safety of the plant and plant site, and must be seen as a compendium of NRS. The FSAR integrates both the licensing requirements and the analytical techniques. The analytical techniques can be applied by using a realistic approach, addressing the uncertainties of the results. This work aims to show an overview of the main analytical techniques that can be applied with a Best Estimated Plus Uncertainty (BEPU) methodology, which is 'the best one can do', as well as the ALARA (As Low As Reasonably Achievable) principle. Moreover, the paper intends to demonstrate the background of the licensing process through the main licensing requirements. (author)

Using of BEPU methodology in a final safety analysis report

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

The Nuclear Reactor Safety (NRS) has been established since the discovery of nuclear fission, and the occurrence of accidents in Nuclear Power Plants worldwide has contributed for its improvement. The Final Safety Analysis Report (FSAR) must contain complete information concerning safety of the plant and plant site, and must be seen as a compendium of NRS. The FSAR integrates both the licensing requirements and the analytical techniques. The analytical techniques can be applied by using a realistic approach, addressing the uncertainties of the results. This work aims to show an overview of the main analytical techniques that can be applied with a Best Estimated Plus Uncertainty (BEPU) methodology, which is 'the best one can do', as well as the ALARA (As Low As Reasonably Achievable) principle. Moreover, the paper intends to demonstrate the background of the licensing process through the main licensing requirements. (author)

75 FR 61619 - Safety Zone; IJSBA World Finals, Lower Colorado River, Lake Havasu, AZ

Science.gov (United States)

2010-10-06

... Sports Boating Association (IJSBA) World Finals. This temporary safety zone is necessary to provide for... International Jet Sports Boating Association (IJSBA) is sponsoring the IJSBA World Finals. The event will... 13211. Technical Standards The National Technology Transfer and Advancement Act (NTTAA) (15 U.S.C. 272...

Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

International Nuclear Information System (INIS)

Ruebel, Andre; Meleshyn, Artur

2014-08-01

The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

Final disposal in deep boreholes using multiple geological barriers. Digging deeper for safety. Proceedings

Energy Technology Data Exchange (ETDEWEB)

Bracke, Guido; Hurst, Stephanie; Merkel, Broder; Mueller, Birgit; Schilling, Frank

2016-03-15

The proceedings of the workshop on final disposal in deep boreholes using multiple geological barriers - digging deeper for safety include contributions on the following topics: international status and safety requirements; geological and physical barriers; deep drilling - shaft building; technical barriers and emplacement technology for high P/T conditions; recovery (waste retrieval); geochemistry and monitoring.

Safety in decommissioning of research reactors

International Nuclear Information System (INIS)

1986-01-01

This Guide covers the technical and administrative considerations relevant to the nuclear aspects of safety in the decommissioning of reactors, as they apply to the reactor and the reactor site. While the treatment, transport and disposal of radioactive wastes arising from decommissioning are important considerations, these aspects are not specifically covered in this Guide. Likewise, other possible issues in decommissioning (e.g. land use and other environmental issues, industrial safety, financial assurance) which are not directly related to radiological safety are also not considered. Generally, decommissioning will be undertaken after planned final shutdown of the reactor. In some cases a reactor may have to be decommissioned following an unplanned or unexpected event of a series or damaging nature occurring during operation. In these cases special procedures for decommissioning may need to be developed, peculiar to the particular circumstances. This Guide could be used as a basis for the development of these procedures although specific consideration of the circumstances which create the need for them is beyond its scope

Safety considerations for research reactors in extended shutdown

International Nuclear Information System (INIS)

2004-01-01

According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

Some safety considerations of liquid lithium as a fusion breeder material

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.

1986-01-01

Test results and conclusions are presented for the reaction of steam with a high temperature lithium pool and for the reaction of high temperature lithium spray with a nitrogen atmosphere. The reactions are characterized and evaluated in regard to the potential for mobilization of radioactive species associated with the liquid breeder under postulated fusion reactor accident conditions. These evaluations include measured lithium temperature responses, atmosphere temperature and pressure responses, gas consumption and generation, aerosol quantities and particle size characterization, and potentially radioactive species releases. Conclusions are made as to the consequences of these safety considerations for the use of lithium as a fusion reactor breeder material

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

International Nuclear Information System (INIS)

Anon.

1985-01-01

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

Safety considerations in the disposal of disused sealed radioactive sources in borehole facilities

International Nuclear Information System (INIS)

2003-08-01

Sealed radioactive sources are used in medicine, industry and research for a wide range of purposes. They can contain different radionuclides in greatly varying amounts. At the end of their useful lives, they are termed 'disused sources' but their activity levels can still be quite high. They are, for all practical purposes, another type of radioactive waste that needs to be disposed of safely. Disused sealed radioactive sources can represent a significant hazard to people if not managed properly. Many countries have no special facilities for the management or disposal of radioactive waste, as they have no nuclear power programmes requiring such facilities. Even in countries with developed nuclear programmes, disused sealed sources present problems as they often fall outside the common categories of radioactive waste for which disposal options have been identified. As a result, many disused sealed sources are kept in storage. Depending on the nature of the storage arrangements, this situation may represent a high potential risk to workers and to the public. The IAEA has received numerous requests for assistance from Member States faced with the problem of safely managing disused sealed sources. The requests have related to both technical and safety aspects. Particularly urgent requests have involved emergency situations arising from unsafe storage conditions and lost sources. There is therefore an important requirement for the development of safe and cost-effective final disposal solutions. Consequently, a number of activities have been initiated by the IAEA to assist Member States in the management of disused sealed sources. The objective of this report is to address safety issues relevant to the disposal of disused sealed sources, and other limited amounts of radioactive waste, in borehole facilities. It is the first in a series of reports aiming to provide an indication of the present issues related to the use of borehole disposal facilities to safely disposal

Considerations on the Application of the IAEA Safety Requirements for the Design of Nuclear Power Plants

International Nuclear Information System (INIS)

2016-05-01

Revised to take into consideration findings from the Fukushima Daiichi nuclear power plant accident, IAEA Safety Standards Series No. SSR-2/1 (Rev. 1), Safety of Nuclear Power Plants: Design, has introduced some new concepts with respect to the earlier safety standard published in the year 2000. The preparation of SSR-2/1 (Rev. 1) was carried out with constant and intense involvement of IAEA Member States, but some new requirements, because of the novelty of the concepts introduced and the complexity of the issues, are not always interpreted in a unique way. The IAEA is confident that a complete clarification and a full understanding of the new requirements will be available when the supporting safety guides for design and safety assessment of nuclear power plants are prepared. The IAEA expects that the effort devoted to the preparation of this publication, which received input and comments from several Member States and experts, will also facilitate and harmonize the preparation or revision of these supporting standards

Considerations on nuclear reactor passive safety systems

International Nuclear Information System (INIS)

2016-01-01

After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

National report of Brazil. Nuclear Safety Convention

International Nuclear Information System (INIS)

1998-09-01

This document represents the national report prepared as a fulfillment of the brazilian obligations related to the Convention on Nuclear Safety. In chapter 2 some details are given about the existing nuclear installations. Chapter 3 provides details about the legislation and regulations, including the regulatory framework and the regulatory body. Chapter 4 covers general safety considerations as described in articles 10 to 16 of the Convention. Chapter 5 addresses to the safety of the installations during siting, design, construction and operation. Chapter 6 describes planned activities to further enhance nuclear safety. Chapter 7 presents the final remarks related to the degree of compliance with the Convention obligations

The impact of regulations, safety considerations and physical limitations on research progress at maximum biocontainment.

Science.gov (United States)

Shurtleff, Amy C; Garza, Nicole; Lackemeyer, Matthew; Carrion, Ricardo; Griffiths, Anthony; Patterson, Jean; Edwin, Samuel S; Bavari, Sina

2012-12-01

We describe herein, limitations on research at biosafety level 4 (BSL-4) containment laboratories, with regard to biosecurity regulations, safety considerations, research space limitations, and physical constraints in executing experimental procedures. These limitations can severely impact the number of collaborations and size of research projects investigating microbial pathogens of biodefense concern. Acquisition, use, storage, and transfer of biological select agents and toxins (BSAT) are highly regulated due to their potential to pose a severe threat to public health and safety. All federal, state, city, and local regulations must be followed to obtain and maintain registration for the institution to conduct research involving BSAT. These include initial screening and continuous monitoring of personnel, controlled access to containment laboratories, accurate and current BSAT inventory records. Safety considerations are paramount in BSL-4 containment laboratories while considering the types of research tools, workflow and time required for conducting both in vivo and in vitro experiments in limited space. Required use of a positive-pressure encapsulating suit imposes tremendous physical limitations on the researcher. Successful mitigation of these constraints requires additional time, effort, good communication, and creative solutions. Test and evaluation of novel vaccines and therapeutics conducted under good laboratory practice (GLP) conditions for FDA approval are prioritized and frequently share the same physical space with important ongoing basic research studies. The possibilities and limitations of biomedical research involving microbial pathogens of biodefense concern in BSL-4 containment laboratories are explored in this review.

The Impact of Regulations, Safety Considerations and Physical Limitations on Research Progress at Maximum Biocontainment

Directory of Open Access Journals (Sweden)

Jean Patterson

2012-12-01

Full Text Available We describe herein, limitations on research at biosafety level 4 (BSL-4 containment laboratories, with regard to biosecurity regulations, safety considerations, research space limitations, and physical constraints in executing experimental procedures. These limitations can severely impact the number of collaborations and size of research projects investigating microbial pathogens of biodefense concern. Acquisition, use, storage, and transfer of biological select agents and toxins (BSAT are highly regulated due to their potential to pose a severe threat to public health and safety. All federal, state, city, and local regulations must be followed to obtain and maintain registration for the institution to conduct research involving BSAT. These include initial screening and continuous monitoring of personnel, controlled access to containment laboratories, accurate and current BSAT inventory records. Safety considerations are paramount in BSL-4 containment laboratories while considering the types of research tools, workflow and time required for conducting both in vivo and in vitro experiments in limited space. Required use of a positive-pressure encapsulating suit imposes tremendous physical limitations on the researcher. Successful mitigation of these constraints requires additional time, effort, good communication, and creative solutions. Test and evaluation of novel vaccines and therapeutics conducted under good laboratory practice (GLP conditions for FDA approval are prioritized and frequently share the same physical space with important ongoing basic research studies. The possibilities and limitations of biomedical research involving microbial pathogens of biodefense concern in BSL-4 containment laboratories are explored in this review.

The Impact of Regulations, Safety Considerations and Physical Limitations on Research Progress at Maximum Biocontainment

Science.gov (United States)

Shurtleff, Amy C.; Garza, Nicole; Lackemeyer, Matthew; Carrion, Ricardo; Griffiths, Anthony; Patterson, Jean; Edwin, Samuel S.; Bavari, Sina

2012-01-01

We describe herein, limitations on research at biosafety level 4 (BSL-4) containment laboratories, with regard to biosecurity regulations, safety considerations, research space limitations, and physical constraints in executing experimental procedures. These limitations can severely impact the number of collaborations and size of research projects investigating microbial pathogens of biodefense concern. Acquisition, use, storage, and transfer of biological select agents and toxins (BSAT) are highly regulated due to their potential to pose a severe threat to public health and safety. All federal, state, city, and local regulations must be followed to obtain and maintain registration for the institution to conduct research involving BSAT. These include initial screening and continuous monitoring of personnel, controlled access to containment laboratories, accurate and current BSAT inventory records. Safety considerations are paramount in BSL-4 containment laboratories while considering the types of research tools, workflow and time required for conducting both in vivo and in vitro experiments in limited space. Required use of a positive-pressure encapsulating suit imposes tremendous physical limitations on the researcher. Successful mitigation of these constraints requires additional time, effort, good communication, and creative solutions. Test and evaluation of novel vaccines and therapeutics conducted under good laboratory practice (GLP) conditions for FDA approval are prioritized and frequently share the same physical space with important ongoing basic research studies. The possibilities and limitations of biomedical research involving microbial pathogens of biodefense concern in BSL-4 containment laboratories are explored in this review. PMID:23342380

Safety considerations for graphene: lessons learnt from carbon nanotubes.

Science.gov (United States)

Bussy, Cyrill; Ali-Boucetta, Hanene; Kostarelos, Kostas

2013-03-19

carbon nanotubes are rare, making comparative considerations of their overall safety and risk assessment challenging. In this Account, we attempt to offer a set of rules for the development of graphene and its derivatives to enhance their overall safety and minimize the risks for adverse reactions in humans from exposure. These rules are: (1) to use small, individual graphene sheets that macrophages in the body can efficiently internalize and remove from the site of deposition; (2) to use hydrophilic, stable, colloidal dispersions of graphene sheets to minimize aggregation in vivo; and (3) to use excretable graphene material or chemically-modified graphene that can be degraded effectively. Such rules can only act as guidelines at this early stage in the development of graphene-based technologies, yet they offer a set of design principles for the fabrication and safe use of graphene material that will come in contact with the human body. In a broader context, the safety risks associated with graphene materials will be entirely dependent on the specific types of graphene materials and how they are investigated or applied. Therefore, generalizations about the toxicity of "graphene" as a whole will be inaccurate, possibly misleading, and should be avoided.

Nuclear safety: an international approach: the convention on nuclear safety

International Nuclear Information System (INIS)

Rosen, M.

1994-01-01

This paper is a general presentation of the IAEA Convention on Nuclear Safety which has already be signed by 50 countries and which is the first legal instrument that directly addresses the safety of nuclear power plants worldwide. The paper gives a review of its development and some key provisions for a better understanding of how this agreement will operate in practice. The Convention consists of an introductory preamble and four chapters consisting of 35 articles dealing with: the principal objectives, definitions and scope of application; the various obligations (general provisions, legislation, responsibility and regulation, general safety considerations taking into account: the financial and human resources, the human factors, the quality assurance, the assessment and verification of safety, the radiation protection and the emergency preparedness; the safety of installations: sitting, design and construction, operation); the periodic meetings of the contracting parties to review national reports on the measures taken to implement each of the obligations, and the final clauses and other judicial provisions common to international agreements. (J.S.). 1 append

Patient safety competence for final-year health professional students: Perceptions of effectiveness of an interprofessional education course.

Science.gov (United States)

Hwang, Jee-In; Yoon, Tai-Young; Jin, Hyeon-Jeong; Park, Yikyun; Park, Ju-Young; Lee, Beom-Joon

2016-11-01

As final-year medical and nursing students will soon play key roles in frontline patient care, their preparedness for safe, reliable care provision is of special importance. We assessed patient safety competencies of final-year health profession students, and the effect of a 1-day patient safety education programme on these competencies. A cross-sectional survey was conducted with 233 students in three colleges of medicine, nursing, and traditional medicine in Seoul. A before-and-after study followed to evaluate the effectiveness of the curriculum. Patient safety competency was measured using the Health-Professional Education for Patients Safety Survey (H-PEPSS) and an objective patient safety knowledge test. The mean scores were 3.4 and 1.7 out of 5.0, respectively. The communication domain was rated the highest and the teamwork domain was rated the lowest. H-PEPSS scores significantly differed between the students from three colleges. The 1-day patient safety education curriculum significantly improved H-PEPSS and knowledge test scores. These results indicated that strengthening patient safety competencies, especially teamwork, of students is required in undergraduate healthcare curricula. A 1-day interprofessional patient safety education programme may be a promising strategy. The findings suggest that interprofessional patient safety education needs to be implemented as a core undergraduate course to improve students' safety competence.

SU-F-P-08: Medical Physics Perspective On Radiation Therapy Quality and Safety Considerations in Low Income Settings

Energy Technology Data Exchange (ETDEWEB)

Van Dyk, J [Western University London, ON (Canada); Meghzifene, A [International Atomic Energy Agency, Vienna (Austria)

2016-06-15

Purpose: The last few years have seen a significant growth of interest in the global radiation therapy crisis. Various organizations are quantifying the need and providing aid in support of addressing the shortfall existing in many low-to-middle income countries (LMICs). The Lancet Oncology Commission report (Lancet Oncol. Sep;16(10):1153-86, 2015) projects a need of 22,000 new medical physicists in LMICs by 2035 if there is to be equal access globally. With the tremendous demand for new facilities, equipment and personnel, it is very important to recognize quality and safety considerations and to address them directly. Methods: A detailed examination of quality and safety publications was undertaken. A paper by Dunscombe (Front. Oncol. 2: 129, 2012) reviewed the recommendations of 7 authoritative reports on safety in radiation therapy and found the 12 most cited recommendations, summarized in order of most to least cited: training, staffing, documentation/standard operating procedures, incident learning, communication/questioning, check lists, QC/PM, dosimetric audit, accreditation, minimizing interruptions, prospective risk assessment, and safety culture. However, these authoritative reports were generally based on input from high income contexts. In this work, the recommendations were analyzed with a special emphasis on issues that are significant in LMICs. Results: The review indicated that there are significant challenges in LMICs with training and staffing ranking at the top in terms quality and safety. Conclusion: With the recognized need for expanding global access to radiation therapy, especially in LMICs, and the backing by multiple support organizations, quality and safety considerations must be overtly addressed. While multidimensional, training and staffing are top priorities. The use of outdated systems with poor interconnectivity, coupled with a lack of systematic QA in high patient load settings are additional concerns. Any support provided to lower

SU-F-P-08: Medical Physics Perspective On Radiation Therapy Quality and Safety Considerations in Low Income Settings

International Nuclear Information System (INIS)

Van Dyk, J; Meghzifene, A

2016-01-01

Purpose: The last few years have seen a significant growth of interest in the global radiation therapy crisis. Various organizations are quantifying the need and providing aid in support of addressing the shortfall existing in many low-to-middle income countries (LMICs). The Lancet Oncology Commission report (Lancet Oncol. Sep;16(10):1153-86, 2015) projects a need of 22,000 new medical physicists in LMICs by 2035 if there is to be equal access globally. With the tremendous demand for new facilities, equipment and personnel, it is very important to recognize quality and safety considerations and to address them directly. Methods: A detailed examination of quality and safety publications was undertaken. A paper by Dunscombe (Front. Oncol. 2: 129, 2012) reviewed the recommendations of 7 authoritative reports on safety in radiation therapy and found the 12 most cited recommendations, summarized in order of most to least cited: training, staffing, documentation/standard operating procedures, incident learning, communication/questioning, check lists, QC/PM, dosimetric audit, accreditation, minimizing interruptions, prospective risk assessment, and safety culture. However, these authoritative reports were generally based on input from high income contexts. In this work, the recommendations were analyzed with a special emphasis on issues that are significant in LMICs. Results: The review indicated that there are significant challenges in LMICs with training and staffing ranking at the top in terms quality and safety. Conclusion: With the recognized need for expanding global access to radiation therapy, especially in LMICs, and the backing by multiple support organizations, quality and safety considerations must be overtly addressed. While multidimensional, training and staffing are top priorities. The use of outdated systems with poor interconnectivity, coupled with a lack of systematic QA in high patient load settings are additional concerns. Any support provided to lower

The project ANSICHT. Safety and demonstration methodology for a final repository in clay formations in Germany; Projekt ANSICHT. Sicherheits- und Nachweismethodik fuer ein Endlager im Tongestein in Deutschland. Synthesebericht

Energy Technology Data Exchange (ETDEWEB)

Jobmann, Michael; Bebiolka, Anke; Jahn, Steffen; and others

2017-03-30

Based on the status of science and technology and under consideration of international repository concepts the fundamental methodology for safety demonstration for a high-level radioactive waste final repository in clay formations Germany was developed. Basic elements of the safety concept are the geological site description and the geo-scientific long-term prognosis on future performance. Another important section is the closure and sealing concept for the mine shafts. In the frame of the project the fundamental elements were developed and documented for model regions in northern and southern Germany. Three independent safety proofs have to be performed: the demonstration of the geological barrier integrity (clay), the demonstration of the geo-technical barrier system integrity - i.e. closure constructions and backfilling of the shafts, and the radiological demonstration that the radionuclide release in the area is lower than the respective limiting value.

Probabilistic safety assessment model in consideration of human factors based on object-oriented bayesian networks

International Nuclear Information System (INIS)

Zhou Zhongbao; Zhou Jinglun; Sun Quan

2007-01-01

Effect of Human factors on system safety is increasingly serious, which is often ignored in traditional probabilistic safety assessment methods however. A new probabilistic safety assessment model based on object-oriented Bayesian networks is proposed in this paper. Human factors are integrated into the existed event sequence diagrams. Then the classes of the object-oriented Bayesian networks are constructed which are converted to latent Bayesian networks for inference. Finally, the inference results are integrated into event sequence diagrams for probabilistic safety assessment. The new method is applied to the accident of loss of coolant in a nuclear power plant. the results show that the model is not only applicable to real-time situation assessment, but also applicable to situation assessment based certain amount of information. The modeling complexity is kept down and the new method is appropriate to large complex systems due to the thoughts of object-oriented. (authors)

Safety program considerations for space nuclear reactor systems

International Nuclear Information System (INIS)

Cropp, L.O.

1984-08-01

This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given

Final report-passive safety optimization in liquid sodium-cooled reactors

International Nuclear Information System (INIS)

Cahalana, J. E.; Hahn, D.

2007-01-01

Implications of Advanced Technology Power Conversion and Design Innovations and Simplifications: Investigations of supercritical CO 2 gas turbine Brayton cycles coupled to the sodium-cooled reactors and innovative concepts for sodium-to-CO 2 heat exchangers were performed to discover new designs for high efficiency electricity production. The objective of the analyses was to characterize the design and safety performance of equipment needed to implement the new power cycle. The project included considerations of heat transfer and power conversion systems arrangements and evaluations of systems performance. Task 4--Post Accident Heat Removal and In-Vessel Retention: Test plans were developed to evaluate (1) freezing and plugging of molten metallic fuel in subassembly geometry, (2) retention of metallic fuel core melt debris within reactor vessel structures, and (3) consequences of intermixing of high pressure CO 2 and sodium. The objective of the test plan development was to provide planning for measurements of data needed to characterize the consequences of very low probability accident sequences unique to metallic fuel and CO 2 Brayton power cycles. The project produced three test plans ready for execution

Project Guarantee 1985. Final repository for high-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

Final disposal of radioactive waste involves preventing the waste from returning from the repository location into the biosphere by means of successively arranged containment measures known as safety barriers. In the present volume NGB 85-04 of the series of reports for Project 'Guarantee' 1985, the safety barrier system for the type C repository for high-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). Safety barriers take the form of both technically constructed containment measures and the siting of the repository in suitable geological formations. The technical safety barrier system in the case of high-level waste comprises: the waste solidification matrix (borosilicate glass), massive steel canisters, encasement of the waste canisters, encasement of the waste canisters in highly compacted bentonite, sealing of vacant storage space and access routes on repository closure. The natural geological safety barriers - the host rock and overlying formations provide sufficiently long deep groundwater flow times from the repository location to the earth's surface and for additional lengthening of radionuclide migration times by means of various chemical and physical retardation mechanisms. The stability of the geological formations is so great that hydrogeological system is protected for a sufficient length of time from deterioration caused, in particular, by erosion. Observations in the final section of the report indicate that input data for the type C repository safety

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

Final Hazard Classification and Auditable Safety Analysis for the N Basin Segment

International Nuclear Information System (INIS)

Kloster, G.L.

1998-08-01

The purposes of this report are to serve as the auditable safety analysis (ASA) for the N Basin Segment, during surveillance and maintenance preceding decontamination and decommissioning; to determine and document the final hazard classification (FHC) for the N Basin Segment. The result of the ASA evaluation are: based on hazard analyses and the evaluation of accidents, no activity could credibly result in an unacceptable exposure to an individual; controls are identified that serve to protect worker health and safety. The results of the FHC evaluation are: potential exposure is much below 10 rem (0.46 rem), and the FHC for the N Basin Segment is Radiological

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Consideration on safety assessment methodologies applied to the near surface repository Baita Bihor

International Nuclear Information System (INIS)

Dogaru, D.

2003-01-01

The Romanian legislation in respect of RAW management is described. The waste facilities in the country are: for low and intermediate level waste - Radioactive Waste Treatment Plant - Bucharest Magurele; Radioactive Waste Treatment Plant - Pitesti; National Repository for Radioactive Waste - Baita Bihor. for spent fuel - Intermediate dry spent fuel storage facility (DICA) - CNE Cernavoda; Intermediate wet spent fuel storage facility WWR-S - Bucharest Magurele. A detailed description of the facilities and waste characterisation are given in the report. Due o insufficient and incomplete information about site characterisation and inventory a Phare project 'Preliminary Safety Analysis for the Low-Level Radioactive Waste Repository Baita Bihor, Romania' has been approved. The project purposes are: to achieve a database with specific parameters; validation of scenarios and conceptual models for normal and altered evolution of the disposal site; validation and qualification of existing calculation methods and identification of the complementary suitable computer codes to be installed in Romania; validation and analyses of the final results expertise PSAR final results; recommendation for further completion of Integrated Performance Assessment. The results, conclusions and recommendations of the project will be included in the Preliminary Safety Analyses Report to be sent to the Romanian Authority - CNCAN for licensing of the repository operation

Safety Considerations in the Chemical Process Industries

Science.gov (United States)

Englund, Stanley M.

There is an increased emphasis on chemical process safety as a result of highly publicized accidents. Public awareness of these accidents has provided a driving force for industry to improve its safety record. There has been an increasing amount of government regulation.

Probabilistic Safety Goals for Nuclear Power Plants; Phases 2-4 / Final Report

International Nuclear Information System (INIS)

Bengtsson, Lisa; Knochenhauer, Michael; Holmberg, Jan-Erik; Rossi, Jukka

2011-05-01

identified, such as the problem of consistency in judgement, comparability of safety goals used in different industries, the relationship between criteria on different levels, and relations between criteria for level 2 and 3 PSA. In parallel, additional context information has been provided. This was achieved by extending the international overview by contributing to and benefiting from a survey on PSA safety criteria which was initiated in 2006 within the OECD/NEA Working Group Risk. Finally, a separate report has been issued providing general guidance concerning the formulation, application and interpretation of probabilistic criteria. 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

Transuranic-contaminated solid waste Treatment Development Facility. Final safety analysis report

International Nuclear Information System (INIS)

Warner, C.L.

1979-07-01

The Final Safety Analysis Report (FSAR) for the Transuranic-Contaminated Solid-Waste Treatment Facility has been prepared in compliance with the Department of Energy (DOE) Manual Chapter 0531, Safety of Nonreactor Nuclear Facilities. The Treatment Development Facility (TDF) at the Los Alamos Scientific Laboratory is a research and development facility dedicated to the study of radioactive-waste-management processes. This analysis addresses site assessment, facility design and construction, and the design and operating characteristics of the first study process, controlled air incineration and aqueous scrub off-gas treatment with respect to both normal and accident conditions. The credible accidents having potentially serious consequences relative to the operation of the facility and the first process have been analyzed and the consequences of each postulated credible accident are presented. Descriptions of the control systems, engineered safeguards, and administrative and operational features designed to prevent or mitigate the consequences of such accidents are presented. The essential features of the operating and emergency procedures, environmental protection and monitoring programs, as well as the health and safety, quality assurance, and employee training programs are described

Transuranic-contaminated solid waste Treatment Development Facility. Final safety analysis report

Energy Technology Data Exchange (ETDEWEB)

Warner, C.L. (comp.)

1979-07-01

The Final Safety Analysis Report (FSAR) for the Transuranic-Contaminated Solid-Waste Treatment Facility has been prepared in compliance with the Department of Energy (DOE) Manual Chapter 0531, Safety of Nonreactor Nuclear Facilities. The Treatment Development Facility (TDF) at the Los Alamos Scientific Laboratory is a research and development facility dedicated to the study of radioactive-waste-management processes. This analysis addresses site assessment, facility design and construction, and the design and operating characteristics of the first study process, controlled air incineration and aqueous scrub off-gas treatment with respect to both normal and accident conditions. The credible accidents having potentially serious consequences relative to the operation of the facility and the first process have been analyzed and the consequences of each postulated credible accident are presented. Descriptions of the control systems, engineered safeguards, and administrative and operational features designed to prevent or mitigate the consequences of such accidents are presented. The essential features of the operating and emergency procedures, environmental protection and monitoring programs, as well as the health and safety, quality assurance, and employee training programs are described.

Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

Energy Technology Data Exchange (ETDEWEB)

Lowe, Andrew; Hayward, Brent (Dedale Asia Pacific, Albert Park VIC 3206 (Australia))

2006-08-15

results of the Safety Culture Perceptions Questionnaire conducted with site managers to access their opinions about the adequacy of the local safety culture; - a framework of safety-related competencies for managers, representing desirable actions for leading and promoting a positive safety culture; - results of an evaluation survey completed by participants at the conclusion of the Management Workshops to assess the utility of this activity. Section 4 of the report, Recommendations for Future Action, highlights nine proposed activities that could be undertaken to build on the outcomes from this project, to support the enhancement of safety culture within the Swedish nuclear industry in the longer term. Specifically, these recommendations propose actions to: 1. Introduce formal processes to ensure the ongoing development of safety related competencies amongst industry managers. 2. Strengthen the resources, contribution, value and profile of Man Technology Organisation (MTO) expertise within nuclear industry sites, in order to promote a better understanding of human performance issues, enhance error management and accident prevention capabilities. 3. Identify ways to embed existing positive safety culture attributes, in an environment of considerable workforce changes resulting from increasing use of contractors and (expected) retirements amongst an ageing industry employee population. 4. Standardise and improve aspects of incident and accident investigation processes and analysis methodologies currently used, to improve information sharing and optimise learning. 5. Implement harmonised MTO / human factors awareness training programs at appropriate levels for all nuclear industry personnel. 6. Formalise the application of applied teamwork training (as per the principles of Crew Resource Management training in aviation) for NPP Control Room Operators, Maintenance workers and other employees working in safety-critical teams. 7. Increase the use of simulation training to

Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume VI. Safety and environmental considerations for licensing

International Nuclear Information System (INIS)

1980-06-01

This volume of the Nonproliferation Alternative Systems Assessment Program report addresses safety and environmental considerations in licensing the principal alternative nuclear reactors and fuel cycles in the United States for large-scale commercial nuclear power plants. In addition, this volume examines the safety and environmental considerations for licensing fuel service centers. These centers, which have been proposed for controlling sensitive fuel-cycle facilities and special nuclear materials, would contain a combination of such facilities as reprocessing plants, fabrication plants, and reactors. For this analysis, two fuel service center concepts were selected - one with power - generating capability and one without

TIBER II/ETR final design report: Volume 3, 5.0 Radiation safety and environment; 6.0 Physics and technology R and D needs

International Nuclear Information System (INIS)

Lee, J.D.

1987-09-01

This paper discusses the design of the TIBER II Tokamak. This particular volume discusses: safety and environmental requirements and design targets; accident analyses; personnel safety and maintenance exposure; effluent control; waste management and decommissioning; safety considerations in building design; and safety and environmental conclusions and recommendations

Consequences of the safety requirement retrieval on existing disposal concepts and requirements to new concepts. Final report; Auswirkungen der Sicherheitsanforderungen Rueckholbarkeit auf existierende Einlagerungskonzepte und Anforderungen an neue Konzepte. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Bollingerfehr, W.; Herold, P.; Doerr, S.; Filbert, W. [DBE Technology GmbH, Peine (Germany)

2014-02-15

In summer 2013 the German Federal Parliament (Bundestag) has restarted the search or a final repository site for high-level radioactive wastes by law. The study is concerned with the consequences of retrieval as additional safety requirement for disposal concepts. Based to the actual knowledge the study concludes that for all possible host rocks (salt, clay and crystalline rock) retrieval is technically possible. Considerable planning effort will be necessary to allow the demonstration of technical feasibility and the reliability of retrieval concepts.

Chronic beryllium disease prevention program; worker safety and health program. Final rule.

Science.gov (United States)

2006-02-09

The Department of Energy (DOE) is today publishing a final rule to implement the statutory mandate of section 3173 of the Bob Stump National Defense Authorization Act (NDAA) for Fiscal Year 2003 to establish worker safety and health regulations to govern contractor activities at DOE sites. This program codifies and enhances the worker protection program in operation when the NDAA was enacted.

Consideration of environmental change in the safety evaluation: Long-term climate scenarios in the Iberian Peninsula

International Nuclear Information System (INIS)

Recreo Jimenez, F.; Ruiz Rivas, C.

1997-01-01

The main objective of this report is twofold. On the one hand, to define the most likely sequences of climate states in the Iberian Peninsula for a period of 125 Ka into the future, to the next interglacial stage, 125 Ka AP; on the other hand, to establish potential climate scenarios during such a period of time determining also the variability ranges of primary climate and climate-related variables of interest to the post-closure performance assessment and underground repository safety evaluations. The report reviews the potential effects of environmental changes on the performance of underground radioactive waste repositories, emphasizing the consideration given to long-term climatic changes in radioactive waste disposal system safety evaluations. (Author)

Considerations on Applying the Method for Assessing the Level of Safety at Work

Directory of Open Access Journals (Sweden)

Costica Bejinariu

2017-07-01

Full Text Available The application of the method for assessing the level of safety at work starts with a document that contains the cover page, the description of the company (name, location, core business, organizational chart etc., description of the work system, a detailed list of its components, and a brief description of the assessment method. It continues with a Microsoft Excel document, which represents the actual application of the method and, finally, there is another document presenting conclusions, proposals, and prioritizations, which leads to the execution of the Prevention and Protection Plan. The present paper approaches the issue of developing the Microsoft Excel document, an essential part of the method for assessing the level of safety at work. The document is divided into a variable number of worksheets, showing the risk categories of general, specific, and management.

Strategies for reactor safety. Final report

Energy Technology Data Exchange (ETDEWEB)

Andersson, K

1997-12-01

The NKS/RAK-1 project formed part of a four-year nuclear research program (1994-1997) in the Nordic countries, the NKS Programme. The project aims were to investigate and evaluate the safety work, to increase realism and reliability of the safety analysis, and to give ideas for how safety can be improved in selected areas. An evaluation of the safety work in nuclear installations in Finland and Sweden was made, and a special effort was devoted to plant modernisation and to see how modern safety standards can be met up with. A combination of more resources and higher efficiency is recommended to meet requirements from plant modernisation and plant renovations. Both the utilities and the safety authorities are recommended to actively follow the evolving safety standards for new reactors. Various approaches to estimating LOCA frequencies have been explored. In particular, a probabilistic model for pipe ruptures due to intergranular stress corrosion has been developed. A survey has been done over methodologies for integrated sequence analysis (ISA), and different approaches have been developed and tested on four sequences. Structured frameworks for integration between PSA and behavioural sciences have been developed, which e.g. have improved PSA. The status of maintenance strategies in Finland and Sweden has been studied and a new maintenance data information system has been developed. (au) 41 refs.

Strategies for reactor safety. Final report

International Nuclear Information System (INIS)

Andersson, K.

1997-12-01

The NKS/RAK-1 project formed part of a four-year nuclear research program (1994-1997) in the Nordic countries, the NKS Programme. The project aims were to investigate and evaluate the safety work, to increase realism and reliability of the safety analysis, and to give ideas for how safety can be improved in selected areas. An evaluation of the safety work in nuclear installations in Finland and Sweden was made, and a special effort was devoted to plant modernisation and to see how modern safety standards can be met up with. A combination of more resources and higher efficiency is recommended to meet requirements from plant modernisation and plant renovations. Both the utilities and the safety authorities are recommended to actively follow the evolving safety standards for new reactors. Various approaches to estimating LOCA frequencies have been explored. In particular, a probabilistic model for pipe ruptures due to intergranular stress corrosion has been developed. A survey has been done over methodologies for integrated sequence analysis (ISA), and different approaches have been developed and tested on four sequences. Structured frameworks for integration between PSA and behavioural sciences have been developed, which e.g. have improved PSA. The status of maintenance strategies in Finland and Sweden has been studied and a new maintenance data information system has been developed. (au)

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-05-15

safety 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. The project provides a comprehensive state-of-the-art description and has contributed to clarifying the history of safety goals both nationally and internationally, the concepts involved in defining and applying probabilistic safety criteria, and the international status and trends in general. It has identified critical issues and the main problem areas. Finally, the project provides useful recommendations and guidance on the definition and application of criteria. Furthermore, the project makes it possible to define criteria stringently, improving the possibilities of argumentation on safety. Generally, this supports efficient use of criteria, yielding more useful PSA results. In this connection, the introduction of ALARP type criteria is judged to provide a very useful way of balancing stringency with the necessary flexibility. There is a possibility of making more active use of lower level criteria. This makes the connection to defence in depth more evident, and opens the perspective of increased control of defence in depth by use of probabilistic methods, including the use as design tools. There is an opportunity for comparison of risk of different NPPs, as well as of comparison of NPP risk with other risks in society. This is judged to provide an opportunity for improved communication on risks with non-PSA experts and with the public in general. However, a necessary condition for meaningful comparisons is to agree on the scope of PSA and methods applied. Obviously, there will also be challenges in the future definition and application of probabilistic safety criteria. These include very general aspects, such as the interpretation of the probability, quality aspects of PSA

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

International Nuclear Information System (INIS)

Bengtsson, L.; Knochenhauer, M.; Holmberg, J.-E.; Rossi, J.

2011-05-01

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. The project provides a comprehensive state-of-the-art description and has contributed to clarifying the history of safety goals both nationally and internationally, the concepts involved in defining and applying probabilistic safety criteria, and the international status and trends in general. It has identified critical issues and the main problem areas. Finally, the project provides useful recommendations and guidance on the definition and application of criteria. Furthermore, the project makes it possible to define criteria stringently, improving the possibilities of argumentation on safety. Generally, this supports efficient use of criteria, yielding more useful PSA results. In this connection, the introduction of ALARP type criteria is judged to provide a very useful way of balancing stringency with the necessary flexibility. There is a possibility of making more active use of lower level criteria. This makes the connection to defence in depth more evident, and opens the perspective of increased control of defence in depth by use of probabilistic methods, including the use as design tools. There is an opportunity for comparison of risk of different NPPs, as well as of comparison of NPP risk with other risks in society. This is judged to provide an opportunity for improved communication on risks with non-PSA experts and with the public in general. However, a necessary condition for meaningful comparisons is to agree on the scope of PSA and methods applied. Obviously, there will also be challenges in the future definition and application of probabilistic safety criteria. These include very general aspects, such as the interpretation of the probability, quality aspects of PSA

Information need about the safety of the final disposal of nuclear waste. Information receiver's views in Eurajoki, Kuhmo and Aeaenekoski municipalities

International Nuclear Information System (INIS)

Hautakangas, H.

1997-03-01

The study analyses the public's information need about the safety issues related to the final disposal of spent nuclear fuel generated by the Finnish nuclear power stations. Locals in three municipalities that are studied as possible sites for final disposal were interviewed for the study. Earlier studies made in Finland had indicated that the public's knowledge about safety issues related to the final disposal was almost opposite to the findings of the natural sciences. Also, the public had expressed a wish to receive more information from the safety authority, the Finnish Centre for Radiation and Nuclear Safety (STUK). This study therefore had two basic objectives: To find out what kind of safety information the locals need and what the safety authority's role could be in providing information. The main results show interest and need especially for information concerning the disposal phases taking place on the ground level, such as nuclear waste transportation and encapsulation. Also, the interviews show a clear need and desire for an impartial actor such as STUK in the information and communication process. (author) (107 refs.)

Investigational new drug safety reporting requirements for human drug and biological products and safety reporting requirements for bioavailability and bioequivalence studies in humans. Final rule.

Science.gov (United States)

2010-09-29

The Food and Drug Administration (FDA) is amending its regulations governing safety reporting requirements for human drug and biological products subject to an investigational new drug application (IND). The final rule codifies the agency's expectations for timely review, evaluation, and submission of relevant and useful safety information and implements internationally harmonized definitions and reporting standards. The revisions will improve the utility of IND safety reports, reduce the number of reports that do not contribute in a meaningful way to the developing safety profile of the drug, expedite FDA's review of critical safety information, better protect human subjects enrolled in clinical trials, subject bioavailability and bioequivalence studies to safety reporting requirements, promote a consistent approach to safety reporting internationally, and enable the agency to better protect and promote public health.

Considerations on monitoring needs of advanced, passive safety light water reactors for severe accident management

International Nuclear Information System (INIS)

Bava, G.; Zambardi, F.

1992-01-01

This paper deals with problems concerning information and related instrumentation needs for Accident Management (AM), with special emphasis on Severe Accidents (SA) in the new advanced, passive safety Light Water Reactors (PLWR), presently in a development stage. The passive safety conception adopted in the plants concerned goes parallel with a deeper consideration of SA, that reflects the need of increasing the plant resistance against conditions going beyond traditional ''design basis accidents''. Further, the role of Accident Management (AM) is still emphasized as last step of the defence in depth concept, in spite of the design efforts aimed to reduce human factor importance; as a consequence, the availability of pertinent information on actual plant conditions remains a necessary premise for performing preplanned actions. This information is essential to assess the evolution of the accident scenarios, to monitor the performances of the safety systems, to evaluate the ultimate challenge to the plant safety, and to implement the emergency operating procedures and the emergency plans. Based on these general purposes, the impact of the new conception on the monitoring structure is discussed, furthermore reference is made to the accident monitoring criteria applied in current plants to evaluate the requirements for possible solutions. (orig.)

Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

International Nuclear Information System (INIS)

Lowe, Andrew; Hayward, Brent

2006-08-01

Safety Culture Perceptions Questionnaire conducted with site managers to access their opinions about the adequacy of the local safety culture; - a framework of safety-related competencies for managers, representing desirable actions for leading and promoting a positive safety culture; - results of an evaluation survey completed by participants at the conclusion of the Management Workshops to assess the utility of this activity. Section 4 of the report, Recommendations for Future Action, highlights nine proposed activities that could be undertaken to build on the outcomes from this project, to support the enhancement of safety culture within the Swedish nuclear industry in the longer term. Specifically, these recommendations propose actions to: 1. Introduce formal processes to ensure the ongoing development of safety related competencies amongst industry managers. 2. Strengthen the resources, contribution, value and profile of Man Technology Organisation (MTO) expertise within nuclear industry sites, in order to promote a better understanding of human performance issues, enhance error management and accident prevention capabilities. 3. Identify ways to embed existing positive safety culture attributes, in an environment of considerable workforce changes resulting from increasing use of contractors and (expected) retirements amongst an ageing industry employee population. 4. Standardise and improve aspects of incident and accident investigation processes and analysis methodologies currently used, to improve information sharing and optimise learning. 5. Implement harmonised MTO / human factors awareness training programs at appropriate levels for all nuclear industry personnel. 6. Formalise the application of applied teamwork training (as per the principles of Crew Resource Management training in aviation) for NPP Control Room Operators, Maintenance workers and other employees working in safety-critical teams. 7. Increase the use of simulation training to enhance non

Manpower analysis in transportation safety. Final report

Energy Technology Data Exchange (ETDEWEB)

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

1977-05-01

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

Safety assessment of geologic repositories for nuclear waste

International Nuclear Information System (INIS)

Bartlett, J.W.; Burkholder, H.C.; Winegardner, W.K.

1977-01-01

Consideration of geologic isolation for final disposition of radioactive wastes has led to the need for evaluation of the safety of the concept. Such evaluations require consideration of factors not encountered in conventional risk analysis: consequences at times and places far removed from the repository site; indirect, complex, and alternative pathways between the waste and the point of potential consequences; a highly limited data base; and limited opportunity for experimental verification of results. R and D programs to provide technical safety evaluations are under way. Three methods are being considered for the probabilistic aspects of the evaluations: fault tree analysis, repository simulation analysis, and system stability analysis. Nuclide transport models, currently in a relatively advanced state of development, are used to evaluate consequences of postulated loss of geologic isolation. This paper outlines the safety assessment methods, unique features of the assessment problem that affect selection of methods and reliability of results, and available results. It also discusses potential directions for future work

Safety culture: personal considerations of an owner/operator

International Nuclear Information System (INIS)

Fuchs, H.

1994-01-01

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

Safety culture in design. Final report

International Nuclear Information System (INIS)

Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M.; Kahlbom, U.; Rollenhagen, C.

2013-04-01

In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

Safety culture in design. Final report

Energy Technology Data Exchange (ETDEWEB)

Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M. [VTT Technical Research Centre of Finland, Espoo (Finland); Kahlbom, U. [Risk Pilot AB, Stockholm (Sweden); Rollenhagen, C. [Vattenfall, Stockholm, (Sweden)

2013-04-15

In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

Dispersion of radioactive material in air and water and consideration of population distribution in site evaluation for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

The IAEA issues Safety Requirements and Safety Guides pertaining to nuclear power plants and activities in the field of nuclear energy, on the basis of its Safety Fundamentals publication on The Safety of Nuclear Installations. The present Safety Guide, which supplements the Code on the Safety of Nuclear Power Plants: Siting, concerns the effects of a nuclear power plant on the surrounding region and the consideration of population distribution in the siting of a plant. This Safety Guide makes recommendations on how to meet the requirements of the Code on the Safety of Nuclear Power Plants: Siting, on the basis of knowledge of the mechanisms for the dispersion of effluents discharged into the atmosphere and into surface water and groundwater. Relevant site characteristics and safety considerations are discussed. Population distribution, the projected population growth rate, particular geographical features, the capabilities of local transport networks and communications networks, industry and agriculture in the region, and recreational and institutional activities in the region should be considered in assessing the feasibility of developing an emergency response plan. In the selection of a site for a facility using radioactive material, such as a nuclear power plant, account should be taken of any local features that might be affected by the facility and of the feasibility of off-site intervention, including emergency response and protective actions. This is in addition to the evaluation of any features of the site itself that might affect the safety of the facility. This Safety Guide recommends methods for the assessment of regional and local characteristics. This Safety Guide supersedes four earlier IAEA Safety Guides, namely: Atmospheric Dispersion in Nuclear Power Plant Siting (Safety Series No. 50-SG-S3 (1980)). Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution (Safety Series No. 50-SG-S4 (1980)). Hydrological

Dispersion of radioactive material in air and water and consideration of population distribution in site evaluation for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

The IAEA issues Safety Requirements and Safety Guides pertaining to nuclear power plants and activities in the field of nuclear energy, on the basis of its Safety Fundamentals publication on The Safety of Nuclear Installations. The present Safety Guide, which supplements the Code on the Safety of Nuclear Power Plants: Siting, concerns the effects of a nuclear power plant on the surrounding region and the consideration of population distribution in the siting of a plant. This Safety Guide makes recommendations on how to meet the requirements of the Code on the Safety of Nuclear Power Plants: Siting, on the basis of knowledge of the mechanisms for the dispersion of effluents discharged into the atmosphere and into surface water and groundwater. Relevant site characteristics and safety considerations are discussed. Population distribution, the projected population growth rate, particular geographical features, the capabilities of local transport networks and communications networks, industry and agriculture in the region, and recreational and institutional activities in the region should be considered in assessing the feasibility of developing an emergency response plan. In the selection of a site for a facility using radioactive material, such as a nuclear power plant, account should be taken of any local features that might be affected by the facility and of the feasibility of off-site intervention, including emergency response and protective actions. This is in addition to the evaluation of any features of the site itself that might affect the safety of the facility. This Safety Guide recommends methods for the assessment of regional and local characteristics. This Safety Guide supersedes four earlier IAEA Safety Guides, namely: Atmospheric Dispersion in Nuclear Power Plant Siting (Safety Series No. 50-SG-S3 (1980)); Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution (Safety Series No. 50-SG-S4 (1980)); Hydrological

Pediatric interventional radiography equipment: safety considerations

International Nuclear Information System (INIS)

Strauss, Keith J.

2006-01-01

This paper discusses pediatric image quality and radiation dose considerations in state-of-the-art fluoroscopic imaging equipment. Although most fluoroscopes are capable of automatically providing good image quality on infants, toddlers, and small children, excessive radiation dose levels can result from design deficiencies of the imaging device or inappropriate configuration of the equipment's capabilities when imaging small body parts. Important design features and setup choices at installation and during the clinical use of the imaging device can improve image quality and reduce radiation exposure levels in pediatric patients. Pediatric radiologists and cardiologists, with the help of medical physicists, need to understand the issues involved in creating good image quality at reasonable pediatric patient doses. The control of radiographic technique factors by the generator of the imaging device must provide a large dynamic range of mAs values per exposure pulse during both fluoroscopy and image recording as a function of patient girth, which is the thickness of the patient in the posterior-anterior projection at the umbilicus (less than 10 cm to greater than 30 cm). The range of pulse widths must be limited to less than 10 ms in children to properly freeze patient motion. Variable rate pulsed fluoroscopy can be leveraged to reduce radiation dose to the patient and improve image quality. Three focal spots with nominal sizes of 0.3 mm to 1 mm are necessary on the pediatric unit. A second, lateral imaging plane might be necessary because of the child's limited tolerance of contrast medium. Spectral and spatial beam shaping can improve image quality while reducing the radiation dose. Finally, the level of entrance exposure to the image receptor of the fluoroscope as a function of operator choices, of added filter thickness, of selected pulse rate, of the selected field-of-view and of the patient girth all must be addressed at installation. (orig.)

Safety considerations of disposal of disused sealed sources in Puspokszilagy Repository, Hungary

International Nuclear Information System (INIS)

2003-01-01

The report presents the management of radioactive waste in Hungary Puspokszilagy Repository (RWTDF) including waste acceptance criteria, safety assessments, Action Plan for the safety improvement and present projects. The Puspokszilagy Repository is a typical near-surface repository, sink into the ground 6 m depth. The facility is a shallow land disposal type, appropriated for disposal of short and medium lived LILW, acceptable for temporary storage of long lived LILW. It consists of vaults containing cells for solidified drummed waste, wells for spent sealed sources, work building for treatment and interim storage and office building for environmental measurements. Two safety assessments have been performed in 2000 and 2002. The new safety assessment confirms the main statements of SA 2000, according to which several waste types can cause serious problems in the distant future: Until the finish of passive control the safety of the environment is guaranteed. After that time it is possible to arise events leading to exceeding of dose restricts (more then 10 mSv/yr but less then 100 mSv/yr), because of disposal of long lived radionuclides (mainly C-14,Tc-99, Ra-226, Th-232, U-234) and significant activities of Cs-137 sources.There are uncertainties in radionuclide amounts and distributions, as well as in the physical and chemical characteristics of the wastes that determine radionuclide mobility and toxicity. The recommendations to improve the safety include: Long lived SSRS in the 'B' and 'D' wells should be removed before the closure of repository. Large Cs-137 sources and long lived sources in the 'A' vaults should be recovered (if its feasible); All vaults should be backfilled to provide chemical conditioning; The waste packaged in plastic bags should be repackaged and compacted into drums or containers; The inventory should be revise. Waste acceptance requirements in the future are: The disposal of long lived radionuclides is no permitted. The long lived waste

Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

International Nuclear Information System (INIS)

Axford, D.J.; Martin, I.C.; McAuley, S.J.

1981-04-01

This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

Safety considerations on LPG storage tanks

International Nuclear Information System (INIS)

Paff, R.

1993-01-01

The safety of liquefied petroleum gas (LPG) storage tanks, in refineries, petrochemicals plants, or distribution storage, is an important concern. Some serious accidents in recent years, have highlighted the need for a good safety policy for such equipment. Accidents in LPG storage are mainly due to losses of containment of the LPG. Formation of a cloud can lead to a ''Unconfined Vapor Cloud Explosion'' (UVCE). Liquid leakage can lead to pool fires in the retention area. In some circumstances the heat input of the tank, combined with the loss of mechanical resistance of the steel under high temperature, can lead to a BLEVE ''Boiling Liquid Expanding Vapor Explosion''. It is obvious that such equipment needs a proper design, maintenance and operating policy. The details to be considered are set out. (4 figures). (Author)

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

Myocardial Contrast Agents – Safety Considerations and Clinical Efficacy in Stress Echocardiography

Directory of Open Access Journals (Sweden)

Maier Anca

2016-11-01

Full Text Available Transthoracic echocardiographic examination is known to be a safe, non-invasive and reproducible method, used in every day clinical practice to obtain important information about cardiac structure and function. Unfortunately, a significant proportion of studies have highlighted the considerable technically difficultly in producing diagnostic images due to a poor acoustic window and more than 33% of patients undergoing stress echocardiography have suboptimal echocardiographic images. All these limitations have led to the use of contrast agents to improve the quality of standard ultrasound examination to provide a better delineation of left ventricle endocardial borders or to obtain information that cannot be achieved by using standard echocardiography, such as assessing myocardial microcirculation and therefore perfusion. This paper sought to review the clinical efficacy and safety of ultrasound contrast agents focusing on stress echocardiography.

Safety considerations of lithium-thionyl chloride cells

Science.gov (United States)

Subbarao, Surampudi; Halpert, Gerald; Stein, Irving

1986-06-01

The use of spirally wound lithium-thionyl chloride (Li-SOCl2) cells is currently limited because of their hazardous behavior. Safety hazards have ranged from mild venting of toxic materials to violent explosions and fires. These incidents may be related to both user- and manufacturer-induced causes. Many explanations have been offered to explain the unsafe behavior of the cells under operating and abuse conditions. Explanations fall into two categories: (1) thermal mechanisms, and (2) chemical mechanisms. However, it is quite difficult to separate the two. Both may be responsible for cell venting or explosion. Some safety problems encountered with these cells also may be due to design deficiencies and ineffective quality control during cell fabrication. A well-coordinated basic and applied research program is needed to develop safe Li-SOCl2 cells. Recommendations include: (1) learnig more about Li-SOL2 cell chemistry; (2) modeling cell and battery behavior; (3) optimizing cell design for safety and performance, (4) implementing quality control procedures; and (5) educating users.

Intergrated plant safety assessment. Systematic evaluation program. Palisades plant, Consumers Power Company, Docket No. 50-255. Final report

International Nuclear Information System (INIS)

1982-10-01

The Nuclear Regulatory Commission (NRC) has published its Final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0820), under the scope of the Systematic Evaluation Program (SEP), for Consumers Power Company's Palisades Plant located in Covert, Van Buren County, Michigan. The SEP was initiated by the NRC to review the design of older operating nuclear reactor plants to reconfirm and document their safety. This report documents the review completed under the SEP for the Palisades Plant. The review has provided for (1) as assessment of the significance of differences between current technical positions on selected safety issues and those that existed when the Palisades Plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when all supplements to the Final IPSAR and the Safety Evaluation Report for converting the license from a provisional to a full-term license have been issued. The report also addresses the comments and recommendations made by the Advisory Committee on Reactor Safeguards in connection with its review of the Draft Report, issued in April 1982

Storage array reflection considerations

International Nuclear Information System (INIS)

Haire, M.J.; Jordan, W.C.; Taylor, R.G.

1997-01-01

The assumptions used for reflection conditions of single containers are fairly well established and consistently applied throughout the industry in nuclear criticality safety evaluations. Containers are usually considered to be either fully water reflected (i.e., surrounded by 6 to 12 in. of water) for safety calculations or reflected by 1 in. of water for nominal (structural material and air) conditions. Tables and figures are usually available for performing comparative evaluations of containers under various loading conditions. Reflection considerations used for evaluating the safety of storage arrays of fissile material are not as well established. When evaluating arrays, it has become more common for analysts to use calculations to demonstrate the safety of the array configuration. In performing these calculations, the analyst has considerable freedom concerning the assumptions made for modeling the reflection of the array. Considerations are given for the physical layout of the array with little or no discussion (or demonstration) of what conditions are bounded by the assumed reflection conditions. For example, an array may be generically evaluated by placing it in a corner of a room in which the opposing walls are far away. Typically, it is believed that complete flooding of the room is incredible, so the array is evaluated for various levels of water mist interspersed among array containers. This paper discusses some assumptions that are made regarding storage array reflection

Denying a patient's final will: public safety vs. medical confidentiality and patient autonomy.

Science.gov (United States)

Gaertner, Jan; Vent, Julia; Greinwald, Ralf; Rothschild, Markus A; Ostgathe, Christoph; Kessel, Rene; Voltz, Raymond

2011-12-01

Especially when caring for patients approaching the end of life, physicians and nursing staff feel committed to fulfilling as many patient desires as possible. However, sometimes a patient's "final will" may threaten public safety. This can lead to severe conflicts, outweighing the physician's obligation and dedication to care for the patient and to respect his autonomy. Yet, public safety can be threatened if confidentiality is not broken. This article provides a concise summary of the medicolegal and ethical fundamentals concerning this difficult situation. If the patient's and others' health and safety are at risk, physicians may (and in some countries must) break medical confidentiality and disclose confidential patient information to the police and other authorities. Physicians should be able to professionally deal with such a conflict in all patients, not only in patients with advanced illness. Copyright © 2011 U.S. Cancer Pain Relief Committee. Published by Elsevier Inc. All rights reserved.

78 FR 16211 - Safety Zone, Corp. Event Finale UHC, St. Thomas Harbor; St. Thomas, U.S. Virgin Islands

Science.gov (United States)

2013-03-14

... 1625-AA00 Safety Zone, Corp. Event Finale UHC, St. Thomas Harbor; St. Thomas, U.S. Virgin Islands... establish a temporary safety zone on the waters of St. Thomas Harbor in St. Thomas, U.S. Virgin Islands... through Friday, except federal holidays. The telephone number is 202-366-9329. See the ``Public...

Regulatory considerations for computational requirements for nuclear criticality safety

International Nuclear Information System (INIS)

Bidinger, G.H.

1995-01-01

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

SAFETY OF PASSIVE HOUSES SUBJECTED TO EARTHQUAKE, FINAL REPORT

Directory of Open Access Journals (Sweden)

Vojko Kilar

2013-12-01

Full Text Available he topic researched within the applied project. "Safety of passive houses subjected to earthquake" stemmed from two otherwise quite unrelated fields, i.e. seismic resistance and energy efficiency that in European countries do not frequently appear together. Just in Slovenia these two fields join each other, so identifying the problem and establishment of research right in Slovenia represents uniqueness and specificity. The majority of Slovenia is situated in area of moderate seismic risk. In order to ensure adequate mechanical resistance and stability of structures constructed in such area, the consideration of seismic effects is required by law. In Slovenia the number of passive houses and energy efficient buildings increases rapidly. However, for the time being the structural solutions that have been developed and broadly applied mainly in the areas with low seismicity (where the structural control to vertical static loads is sufficient are used. In earthquake-prone areas also adequate resistance to dynamic seismic effects have to be assured.

Operational safety and radioprotection considerations when designing the ILW-LL disposal zone

International Nuclear Information System (INIS)

Voinis, S.; Roulet, A.; Claudel, D.; Lesavre, A.

2008-01-01

As for any other nuclear industrial facility, in a radioactive waste repository the various waste disposal operational activities from construction to closure can present a risk to human (workers and public) and the environment. In accordance with the December 30, 1991 French Waste Act, Andra has conducted feasibility studies regarding the disposal of HLW and ILW-LL waste in a clay host formation. The 'Dossier 2005 - Clay' includes a description of the operational safety analysis that was conducted for ILW-LL waste disposal in underground horizontal drifts. The objective of this paper is to present that safety analysis and its impact on the design at the feasibility stage. The safety analysis covered the operations from the reception of the waste transport casks to the disposal of the waste disposal package in its final emplacement location inside the disposal cell. Since the surface facilities' operations are similar to those of other nuclear ones, this paper focuses on the specificity of the deep repository, i.e. the operational safety and radioprotection aspects applied to the deep disposal drift. Andra has selected an ILW-LL design based on large horizontal drifts (diameters of 10 to 12 m, and lengths of 250 m). The primary waste packages are put inside a specific concrete overpack before their disposal. These overpacks are remotely stacked inside the horizontal drifts. The operational safety analysis aims to ensure that risks are kept under control through provisions in the design of the repository and by operating the facility in compliance with operational requirements and the safety functions. The requirements and the safety functions, developed at this stage of the feasibility studies, will be explained. The operational safety analysis is structured around physical components and real activities (construction, operation, closure) through a dedicated risk analysis. Due to the large variety of different ILW-LL waste, in order to identify the potential

Safety considerations of lithium-thionyl chloride cells

Energy Technology Data Exchange (ETDEWEB)

Subbarao, S.; Halpert, G.; Stein, I.

1986-06-01

The use of spirally wound lithium-thionyl chloride (Li-SOCl/sub 2/) cells is currently limited because of their hazardous behavior. Safety hazards have ranged from mild venting of toxic materials to violent explosions and fires. These incidents may be related to both user- and manufacturer-induced causes. Many explanations have been offered to explain the unsafe behavior of the cells under operating and abuse conditions. Explanations fall into two categories: (1) thermal mechanisms, and (2) chemical mechanisms. However, it is quite difficult to separate the two. Both may be responsible for cell venting or explosion. Some safety problems encountered with these cells also may be due to design deficiencies and ineffective quality control during cell fabrication. A well-coordinated basic and applied research program is needed to develop safe Li-SOCl/sub 2/ cells. Recommendations include: (1) learnig more about Li-SOCl/sub 2/ cell chemistry; (2) modeling cell and battery behavior; (3) optimizing cell design for safety and performance, (4) implementing quality control procedures; and (5) educating users.

Ab initio chemical safety assessment: A workflow based on exposure considerations and non-animal methods.

Science.gov (United States)

Berggren, Elisabet; White, Andrew; Ouedraogo, Gladys; Paini, Alicia; Richarz, Andrea-Nicole; Bois, Frederic Y; Exner, Thomas; Leite, Sofia; Grunsven, Leo A van; Worth, Andrew; Mahony, Catherine

2017-11-01

We describe and illustrate a workflow for chemical safety assessment that completely avoids animal testing. The workflow, which was developed within the SEURAT-1 initiative, is designed to be applicable to cosmetic ingredients as well as to other types of chemicals, e.g. active ingredients in plant protection products, biocides or pharmaceuticals. The aim of this work was to develop a workflow to assess chemical safety without relying on any animal testing, but instead constructing a hypothesis based on existing data, in silico modelling, biokinetic considerations and then by targeted non-animal testing. For illustrative purposes, we consider a hypothetical new ingredient x as a new component in a body lotion formulation. The workflow is divided into tiers in which points of departure are established through in vitro testing and in silico prediction, as the basis for estimating a safe external dose in a repeated use scenario. The workflow includes a series of possible exit (decision) points, with increasing levels of confidence, based on the sequential application of the Threshold of Toxicological (TTC) approach, read-across, followed by an "ab initio" assessment, in which chemical safety is determined entirely by new in vitro testing and in vitro to in vivo extrapolation by means of mathematical modelling. We believe that this workflow could be applied as a tool to inform targeted and toxicologically relevant in vitro testing, where necessary, and to gain confidence in safety decision making without the need for animal testing.

Safety requirements of the BMU to be met in final storage of heat-producing waste: An evaluation

International Nuclear Information System (INIS)

Thomauske, Bruno

2009-01-01

On August 12, 2008, The German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) published a draft of July 29, 2008 of the ''Safety Requirements to Be Met in Final Storage of Heat-producing Radioactive Waste.'' As announced by the BMU, these safety requirements are to bring up to the state of the art the safety criteria of 1983. Over a couple of years, efforts had been made to adapt the criteria to the internationally accepted standard as demanded by the Advisory Committees on Reactor Safeguards (RSK) and Radiation Protection (SSK). There is no waste management concept underlying the safety requirements. As a consequence, the draft should be withdrawn by the Federal Ministry for the Environment and replaced by a version revised from scratch and offering assured quality. (orig./GL)

Terrain Safety Assessment in Support of the Mars Science Laboratory Mission

Science.gov (United States)

Kipp, Devin

2012-01-01

In August 2012, the Mars Science Laboratory (MSL) mission will pioneer the next generation of robotic Entry, Descent, and Landing (EDL) systems by delivering the largest and most capable rover to date to the surface of Mars. The process to select the MSL landing site took over five years and began with over 50 initial candidate sites from which four finalist sites were chosen. The four finalist sites were examined in detail to assess overall science merit, EDL safety, and rover traversability on the surface. Ultimately, the engineering assessments demonstrated a high level of safety and robustness at all four finalist sites and differences in the assessment across those sites were small enough that neither EDL safety nor rover traversability considerations could significantly discriminate among the final four sites. Thus the MSL landing site at Gale Crater was selected from among the four finalists primarily on the basis of science considerations.

Ultrasonic imaging: safety considerations

Science.gov (United States)

ter Haar, Gail

2011-01-01

Modern ultrasound imaging for diagnostic purposes has a wide range of applications. It is used in obstetrics to monitor the progress of pregnancy, in oncology to visualize tumours and their response to treatment, and, in cardiology, contrast-enhanced studies are used to investigate heart function and physiology. An increasing use of diagnostic ultrasound is to provide the first photograph for baby's album—in the form of a souvenir or keepsake scan that might be taken as part of a routine investigation, or during a visit to an independent high-street ‘boutique’. It is therefore important to ensure that any benefit accrued from these applications outweighs any accompanying risk, and to evaluate the existing ultrasound bio-effect and epidemiology literature with this in mind. This review considers the existing laboratory and epidemiological evidence about the safety of diagnostic ultrasound and puts it in the context of current clinical usage. PMID:22866238

Ultrasonic imaging: safety considerations.

Science.gov (United States)

Ter Haar, Gail

2011-08-06

Modern ultrasound imaging for diagnostic purposes has a wide range of applications. It is used in obstetrics to monitor the progress of pregnancy, in oncology to visualize tumours and their response to treatment, and, in cardiology, contrast-enhanced studies are used to investigate heart function and physiology. An increasing use of diagnostic ultrasound is to provide the first photograph for baby's album-in the form of a souvenir or keepsake scan that might be taken as part of a routine investigation, or during a visit to an independent high-street 'boutique'. It is therefore important to ensure that any benefit accrued from these applications outweighs any accompanying risk, and to evaluate the existing ultrasound bio-effect and epidemiology literature with this in mind. This review considers the existing laboratory and epidemiological evidence about the safety of diagnostic ultrasound and puts it in the context of current clinical usage.

Health and safety in clinical laboratories in developing countries: safety considerations.

Science.gov (United States)

Ejilemele, A A; Ojule, A C

2004-01-01

Clinical laboratories are potentially hazardous work areas. Health and safety in clinical laboratories is becoming an increasingly important subject as a result of the emergence of highly infectious diseases such as hepatitis and HIV. This is even more so in developing countries where health and safety have traditionally been regarded as low priority issues, considering the more important health problems confronting the health authorities in these countries. We conducted a literature search using the medical subheadings titles on the INTERNET over a period of twenty years and summarized our findings. This article identifies hazards in the laboratories and highlights measures to make the laboratory a safer work place. It also emphasizes the mandatory obligations of employers and employees towards the attainment of acceptable safety standards in clinical laboratories in Third World countries in the face of the current HIV/AIDS epidemic in many of these developing countries especially in the sub-Saharan Africa while accommodating the increasing work load in these laboratories. Both the employer and the employee have major roles to play in the maintenance of a safe working environment. This can be achieved if measures discussed are incorporated into everyday laboratory practice.

Thermal reactor safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

Thermal reactor safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

New safety concept for geological disposal in Japan - -16339

International Nuclear Information System (INIS)

Kitayama, Kazumi

2009-01-01

This paper describes a new safety concept for the Japanese geological disposal program, which is a development of the conventional multi-barrier system concept. The Japanese government established the 'Nuclear Waste Management Organization of Japan' (NUMO) as an implementation body in 2000 based on the 'Final disposal act' following the publication of the 'H-12 Report', which confirmed the scientific and engineering feasibility of HLW geological disposal in Japan. Since then, NUMO has undertaken further technical developments aimed at achieving safe and efficient implementation of final disposal. The safety concept developed in the 'H-12 Report' provides sufficient safety on the basis of site-generic considerations. However, it is considered to be over-conservative and therefore does not represent the most probable performance of the engineered or natural barriers. Recently, concrete measures have been proposed requiring the safety case to be presented in terms of a realistic assessment of the most probable performance. This approach takes into account the safety functions of both engineered and natural barriers as well as the long-term static geochemical equilibrium. In particular, the evolution of the safety performance of engineered and natural barriers can be efficiently augmented by the realistic long-term geochemical equilibrium. (author)

78 FR 16208 - Safety Zone; V. I. Carnival Finale; St. Thomas Harbor; St. Thomas, U.S. Virgin Islands

Science.gov (United States)

2013-03-14

... 1625-AA00 Safety Zone; V. I. Carnival Finale; St. Thomas Harbor; St. Thomas, U.S. Virgin Islands AGENCY... establish a safety zone on the waters of St. Thomas Harbor in St. Thomas, U. S. Virgin Islands during the V... between 9 a.m. and 5 p.m., Monday through Friday, except federal holidays. The telephone number is 202-366...

LMFBR safety criteria: cost-benefit considerations under the constraint of an a priori risk criterion

International Nuclear Information System (INIS)

Hartung, J.

1979-01-01

The role of cost-benefit considerations and a priori risk criteria as determinants of Core Disruptive Accident (CDA)-related safety criteria for large LMFBR's is explored with the aid of quantitative risk and probabilistic analysis methods. A methodology is described which allows a large number of design and siting alternatives to be traded off against each other with the goal of minimizing energy generation costs subject to the constraint of both an a priori risk criterion and a cost-benefit criterion. Application of this methodology to a specific LMFBR design project is described and the results are discussed. 5 refs

Consideration of timescales in the finnish safety regulations for spent fuel disposal

International Nuclear Information System (INIS)

Ruokola, E.

2010-01-01

has taken upon itself the identification of the impacts on the biosphere from the releases of disposed radionuclides. The Farthest Future: At about 250 000 years, the total radioactivity in spent nuclear fuel becomes very close to that in the natural uranium from which the fuel was fabricated. In this time frame, the hazard posed by a spent fuel repository will be comparable to that of a medium sized natural uranium deposit and the repository might be regarded as being part of the nature. Accordingly, for the time period beyond one million years, the regulations do not require any rigorous quantitative safety assessments but the judgement of safety can be based on more qualitative considerations, such as bounding analyses with simplified methods, comparisons with natural analogues and observations of the geological history of the site

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of real component geometries (phase 1). Final report; Erhoehung der Komponentensicherheit durch verbesserte Verfahren zur Eigenspannungsanalyse. Teilvorhaben. Beruecksichtigung realer Komponentengeometrien (Phase 1). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Nau, Andreas; Scholtes, B.

2014-07-24

-axial plane stress states were taken into account in this study. In the case of component's edges minor influences on the calculated stresses could be observed using standard calibration coefficients. Thus, a specific calibration is not considered necessary. The effects on shoulders could be assessed analogously. In the case of the claddings a remarkable influence could only be observed, if their thickness is below 2 mm. Essential benefits were reached through pre-experiments. As a consequence, the experimental execution was improved in terms of the material removal technique. The experimental data preparation was investigated in addition, which lead to the approach not to condition the experimental strains but rather the calculated stresses. This method is important for the experiments of phase 2 within this project. As a consequence of a successful project, the present knowledge base will be considerably improved and will be available for various engineering fields. Especially, the quantitative consideration of real residual stress states for optimized component designs will be feasible and finally the consequences of residual stresses an the component's safety, which are used in nuclear facilities, can be evaluated. The findings of the first numerical and theoretical research period (phase 1) at Kassel University are documented in this report.

Design trade-offs in view of safety considerations

International Nuclear Information System (INIS)

Saji, G.; Kishida, K.; Inoue, T.

1978-01-01

In view of resolving conflicting demands of cost, safety, flexibility of operation and design margins, safety design of various plant systems is discussed referring to their weight on construction costs. An influence of hypothetical core disruptive accident (HCDA) and loss of piping integrity (LOPI) on plant design and thus on construction materials is discussed, in optimising future commercial FBR plants. (author)

Safety balance: Analysis of safety systems

International Nuclear Information System (INIS)

Delage, M.; Giroux, C.

1990-12-01

Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

Decision nearing on final disposal of spent fuel in Finland

International Nuclear Information System (INIS)

Vira, J.

2000-01-01

The programme for final disposal of spent fuel from Finnish nuclear power plants is entering into important phase: in the year 2000 the Finnish Government is expected to decide whether the proposal made by Posiva Oy on the spent fuel disposal is in line with the overall good of society. Associated with the decision is also Posiva's proposal on siting the disposal facility at Olkiluoto in Eurajoki municipality on the western coast of Finland. An important document underlying Posiva's application for this principle decision is the report of the environmental impact assessment, which was completed in 1999. Safety considerations play an important role in the application. New assessments have, therefore, been made on both the operational and long-term safety as well as on safety of spent fuel transportation. (author)

Fifth national report of Brazil for the nuclear safety convention

International Nuclear Information System (INIS)

2010-01-01

This Fifth National Report is a new update to include relevant information for the period of 2007/2009. This document represents the national report prepared as a fulfillment of the Brazilian obligations related to the Convention on Nuclear Safety. In chapter 2 some details are given about the existing nuclear installations. Chapter 3 provides details about the legislation and regulations, including the regulatory framework and the regulatory body. Chapter 4 covers general safety considerations as described in articles 10 to 16 of the Convention. Chapter 5 addresses to the safety of the installations during siting, design, construction and operation. Chapter 6 describes planned activities to further enhance nuclear safety. Chapter 7 presents the final remarks related to the degree of compliance with the Convention obligations

Safety considerations in next step fusion design and beyond

International Nuclear Information System (INIS)

Holland, D.F.

1990-01-01

Recent U.S. and international design studies provide insights into the potential safety and environmental advantages of fusion as well as the development needed to realize this potential. We in the Fusion Safety Program at EG ampersand G Idaho have analyzed the Compact Ignition Tokamak (CIT), the International Thermonuclear Engineering Reactor (ITER), and the Advanced Reactor Innovative Engineering Study (ARIES). I have reviewed these three designs to determine issues related to meeting the safety and the environmental goals that guide fusion development in the U.S. The paper lists safety and environmental issues that are generic to fusion and approaches to favorably resolve each issue. The technical developments that have the highest potential of contributing to improving the safety and environmental attractiveness of fusion are identified and discussed. These developments are in the areas of low-activation materials, plasma- facing components, and plasma physics relating to off-normal plasma events and tritium burn-up. 8 refs., 7 tabs

Investigation and consideration on the framework of oversight-based safety regulation. U.S. NRC 'Risk-Informed, Performance-Based' Regulation

International Nuclear Information System (INIS)

Saji, Gen

2001-01-01

Regulation on safety, environment and health in Japan has before today been intended to correspond with an accident at forms of reinforcement of national standards and monitoring, if any. However, as it was thought that such regulation reinforcement was afraid to bring some social rigidity, and to weaken independent responsibility, as a result, because of anxiety of losing peoples' merits inversely, some fundamental directivity such as respect of self-responsibility principle' and 'necessary and least limit of regulation' were selected as a part of political innovation. On the other hand, at a background of wide improvements on various indexing values showing operation results of nuclear power stations in U.S.A., private independent effort on upgrading of safety is told to largely affect at beginning of INPO (Institute of Nuclear Power Operations), without regulation reinforcement of NRC side. This is a proof of concrete effect of transfer to oversight-based safety regulation. Here were introduced on nuclear safety in U.S.A. at a base of some references obtained on entering the 'MIT summer specialist program. Nuclear system safety', on focussing at new safety regulation of NRC and its effect and so on, and adding some considerations based on some knowledge thereafter. (G.K.)

Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

1990-01-01

The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

White paper on nuclear safety in 2000

International Nuclear Information System (INIS)

2001-04-01

This report is composed of three parts and a subjective part Part 1 includes special articles on the measures for the security of nuclear safety and the future problems described from the beginning of the security. Taking consideration that there exists potential risk in the utilization of nuclear energy in addition to the previous accidents in the area of nuclear energy, future measures to take for safety security were discussed as well as the reorganization of government facilities. In addition, the measures for nuclear safety according to the special nuclear disaster countermeasure law and the future problems were described. In Part 2, the trend of nuclear safety in 2000 and the actual effects of 'the basic principle for the countermeasures of the hour' proposed by the nuclear safety commission were outlined. Moreover, the activities of the commission in 2000 were briefly described. In Part 3, various activities for security of nuclear safety, the safety regulation system and the disaster protection system in nuclear facilities, nuclear safety researches in Japan were described in addition to international cooperation as to nuclear safety. Finally, various materials related to the nuclear safety commission, and the materials on the practical activities for nuclear safety were listed in the subjective part. (M.N.)

Storage array reflection considerations

International Nuclear Information System (INIS)

Haire, M.J.; Jordan, W.C.; Taylor, R.G.

1997-01-01

The assumptions used for reflection conditions of single containers are fairly well established and consistently applied throughout the industry in nuclear criticality safety evaluations. Containers are usually considered to be either fully water-reflected (i.e. surrounded by 6 to 12 in. of water) for safety calculations or reflected by 1 in. of water for nominal (structural material and air) conditions. Tables and figures are usually available for performing comparative evaluations of containers under various loading conditions. Reflection considerations used for evaluating the safety of storage arrays of fissile material are not as well established

Strategies for reactor safety: Preventing loss of coolant accidents. Final report

International Nuclear Information System (INIS)

Lydell, B.O.Y.

1997-12-01

This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; b) In-service inspection practices and how they influence piping reliability; and c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG)

Low-water considerations for NPP sites

International Nuclear Information System (INIS)

Jensen, J.; Frank, T.; Wahl, T.

2009-01-01

In the frame of reactor safety considerations flood protection and the evaluation of low-water events are safety relevant issues. Therefore low-water statistics were performed for the coastal region Cuxhaven and the Elbe river estuary. The consideration of the longitudinal profile of water levels during low-tide in connection with surface water effects, morphodynamic changes of the river and anthropogenic modifications is of importance for conclusions concerning the NPP sites in some distance of the reference tide gauge. The authors performed a statistical low-tide analysis for the NPP sites Brunsbuettel and Brokdorf.

FDA drug safety communications: a narrative review and clinical considerations for older adults.

Science.gov (United States)

Marcum, Zachary A; Vande Griend, Joseph P; Linnebur, Sunny A

2012-08-01

adults. Finally, several warnings were made about statins. Routine periodic monitoring of liver enzymes does not appear to be effective in detecting or preventing serious liver injury from statin use; thus, liver enzymes are no longer recommended to be routinely monitored. Statin-induced cognitive changes are rare, and insufficient evidence is currently available to establish causality. Statins appear to moderately increase the risk of developing diabetes (versus placebo), and regular screening for diabetes should be considered, especially for patients taking high-dose statins and patients with multiple risk factors for diabetes. FDA drug safety communications incorporate complex methodologies that investigate the risks (and relative benefits) of medication therapy. Clinicians caring for older adults need to be aware of the most current evidence behind these drug risks to effectively communicate with and care for their patients. Copyright © 2012 Elsevier HS Journals, Inc. All rights reserved.

Handling final storage of unreprocessed spent nuclear fuel

International Nuclear Information System (INIS)

1978-01-01

The present second report from KBS describes how the safe final storage of spent unreprocessed nuclear fuel can be implemented. According to the Swedish Stipulation Law, the owner must specify in which form the waste is to be stored, how final storage is to be effected, how the waste is to be transported and all other aspects of fuel handling and storage which must be taken into consideration in judging whether the proposed final storage method can be considered to be absolutely safe and feasible. Thus, the description must go beyond general plans and sketches. The description is therefore relatively detailed, even concerning those parts which are less essential for evaluating the safety of the waste storage method. For those parts of the handling chain which are the same for both alternatives of the Stipulation Law, the reader is referred in some cases to the first report. Both of the alternatives of the Stipulation Law may be used in the future. Handling equipment and facilities for the two storage methods are so designed that a combination in the desired proportions is practically feasible. In this first part of the report are presented: premises and data, a description of the various steps of the handling procedure, a summary of dispersal processes and a safety analysis. (author)

Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume VI. Safety and environmental considerations for licensing

International Nuclear Information System (INIS)

1979-12-01

Volume 6 of the Nonproliferation Alternative Systems Assessment Program report addresses safety and environmental considerations in licensing the principal alternative nuclear reactors and fuel cycles in the United States for large-scale commercial nuclear power plants. In addition, this volume examines the safety and environmental considerations for licensing fuel service centers. These centers, which have been proposed for controlling sensitive fuel-cycle facilities and special nuclear materials, would contain a combination of such facilities as reprocessing plants, fabrication plants, and reactors. For this analysis, two fuel service center concepts were selected - one with power-generating capability and one without. This volume also provides estimates of the time required for development of large-scale commercial reactor systems to reach the construction permit application stage and for fuel-cycle facilities to reach the operating license application stage, which is a measure of the relative technical status of alternative nuclear systems

Radiation protection and safety for final disposal of radioactive wastes stored in Abadia de Goias, Brazil

International Nuclear Information System (INIS)

1991-01-01

This standard aims to satisfy the radiation protection and safety conditions required by Brazilian Nuclear Energy Commission (CNEN) for final disposal of radioactive wastes stored in Abadia de Goias. These wastes are products of the accident happened in 1987 caused by the Cs-137 source violation. (M.V.M.)

Organisational factors. Their definition and influence on nuclear safety. Final report

International Nuclear Information System (INIS)

Baumont, G.; Wahlstroem, B.; Sola, R.; Williams, J.; Frischknecht, A.; Wilpert, B.; Rollenhagen, C.

2000-12-01

The importance of organisational factors in the operational safety and efficiency of nuclear power plants (NPP) has been recognised by many organisations around the world. Despite this recognition, however, there are as yet very few methods by which organisational factors can be systematically assessed and improved. The majority of research efforts applied so far have tended to be modest and scattered. The ORFA project was created as a remedy to these problems. The objective of the project is to create a better understanding of how organisation and management factors influence nuclear safety. A key scientific objective of the project is to identify components of a theoretical framework, which would help in understanding the relationships between organisational factors and nuclear safety. Three work packages were planned. First, a review of literature listed out the identified factors and methods for assessing them. Then, a draft version of the present report was prepared to clarify the environment context and the main issues of the topics. This draft was discussed at the ORFA seminar in Madrid 21-22 October 1999. During the seminar views and comments were collected on preliminary results of the project. Finally, this information has been integrated in the present and other reports and will be used to give further guidance to the European Commission in the development of forthcoming research programmes in the field. The project has addressed nuclear safety taking a broad perspective, which reflected and took into account the views of senior NPP management and regulators. The questions discussed during the project have been: how can organisational factors be included in safety assessments, how can good and bad operational practices be identified, which methods can be used for detecting weak signals of deteriorating performance, how should incidents be analysed with respect to organisational factors to give the largest learning benefit, how can data on organisational

Storage of Spent Nuclear Fuel. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

Critical safety issues in the design of fusion machines

International Nuclear Information System (INIS)

Kramer, W.

1991-01-01

In the course of developing fusion machines both general safety considerations and safety assessments for the various components and systems of actual machines increase in number and become more and more coherent. This is particularly true for the NET/ITER projects where safety analysis plays an increasing role for the design of the machine. Since in a D/T tokamak the radiological hazards will be dominant basic radiological safety objectives are discussed. Critical safety issues as identified in particular by the NET/ITER community are reviewed. Subsequently, issues of major concern are considered both for normal operation and for conceivable accidents. The following accidents are considered to be crucial: Loss of cooling in plasma facing components, loss of vacuum, tritium system failure, and magnet system failure. To mitigate accident consequences a confinement concept based on passive features and multiple barriers including detritiation and filtering has to be applied. The reactor building as final barrier needs special attention to cope with both internal and external hazards. (orig.)

Basic considerations for the safety analysis report of the Greek Research Reactor-1 (GRR-1)

International Nuclear Information System (INIS)

Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

1980-09-01

The basic considerations upon which the new revised Safety Analysis Report (SAR) for the GRR-1 will be based are presented. The format and the content the SAR will follow are given. A number of credible and less credible accidents is briefly analysed on the basis of present knowledge and experience for similar reactors, as well as the experience gained in the last 10 years of the GRR-1 operation at 5 MW. The accident caused by partial blockage of the cooling flow is considered to be the Maximum Credible Accident (MCA) for the GRR-1. The MCA is analysed and its radiological impact to the environment is estimated using conservative assumptions. (T.A.)

Fourth national report of Brazil for the nuclear safety convention. Sep. 2007

International Nuclear Information System (INIS)

2007-09-01

This Fourth National Report of Brazil is a new update to include relevant information of the period of 2004-2007. This document represents the national report prepared as a fulfillment of the brazilian obligations related to the Convention on Nuclear Safety. In chapter 2 some details are given about the existing nuclear installations. Chapter 3 provides details about the legislation and regulations, including the regulatory framework and the regulatory body. Chapter 4 covers general safety considerations as described in articles 10 to 16 of the Convention. Chapter 5 addresses to the safety of the installations during siting, design, construction and operation. Chapter 6 describes planned activities to further enhance nuclear safety. Chapter 7 presents the final remarks related to the degree of compliance with the Convention obligations

Fourth national report of Brazil for the nuclear safety convention. Sep. 2007

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-09-15

This Fourth National Report of Brazil is a new update to include relevant information of the period of 2004-2007. This document represents the national report prepared as a fulfillment of the brazilian obligations related to the Convention on Nuclear Safety. In chapter 2 some details are given about the existing nuclear installations. Chapter 3 provides details about the legislation and regulations, including the regulatory framework and the regulatory body. Chapter 4 covers general safety considerations as described in articles 10 to 16 of the Convention. Chapter 5 addresses to the safety of the installations during siting, design, construction and operation. Chapter 6 describes planned activities to further enhance nuclear safety. Chapter 7 presents the final remarks related to the degree of compliance with the Convention obligations

Is road safety management linked to road safety performance?

Science.gov (United States)

Papadimitriou, Eleonora; Yannis, George

2013-10-01

This research aims to explore the relationship between road safety management and road safety performance at country level. For that purpose, an appropriate theoretical framework is selected, namely the 'SUNflower' pyramid, which describes road safety management systems in terms of a five-level hierarchy: (i) structure and culture, (ii) programmes and measures, (iii) 'intermediate' outcomes'--safety performance indicators (SPIs), (iv) final outcomes--fatalities and injuries, and (v) social costs. For each layer of the pyramid, a composite indicator is implemented, on the basis of data for 30 European countries. Especially as regards road safety management indicators, these are estimated on the basis of Categorical Principal Component Analysis upon the responses of a dedicated road safety management questionnaire, jointly created and dispatched by the ETSC/PIN group and the 'DaCoTA' research project. Then, quasi-Poisson models and Beta regression models are developed for linking road safety management indicators and other indicators (i.e. background characteristics, SPIs) with road safety performance. In this context, different indicators of road safety performance are explored: mortality and fatality rates, percentage reduction in fatalities over a given period, a composite indicator of road safety final outcomes, and a composite indicator of 'intermediate' outcomes (SPIs). The results of the analyses suggest that road safety management can be described on the basis of three composite indicators: "vision and strategy", "budget, evaluation and reporting", and "measurement of road user attitudes and behaviours". Moreover, no direct statistical relationship could be established between road safety management indicators and final outcomes. However, a statistical relationship was found between road safety management and 'intermediate' outcomes, which were in turn found to affect 'final' outcomes, confirming the SUNflower approach on the consecutive effect of each layer

Baking soda as an abrasive in toothpastes: Mechanism of action and safety and effectiveness considerations.

Science.gov (United States)

Hara, Anderson T; Turssi, Cecilia P

2017-11-01

Toothpastes can be formulated with different abrasive systems, depending on their intended clinical application. This formulation potentially affects their effectiveness and safety and, therefore, requires proper understanding. In this article, the authors focused on abrasive aspects of toothpastes containing sodium bicarbonate (baking soda), which have gained considerable attention because of their low abrasivity and good compatibility, while providing clinical effectiveness (further detailed in the other articles of this special issue). The authors first appraised the role of toothpaste abrasivity on tooth wear, exploring some underlying processes and the existing methods to determine toothpaste abrasivity. The authors reviewed the available data on the abrasivity of toothpastes containing baking soda and reported a summary of findings highlighting the clinical implications. On the basis of the collected evidence, baking soda has an intrinsic low-abrasive nature because of its comparatively lower hardness in relation to enamel and dentin. Baking soda toothpastes also may contain other ingredients, which can increase their stain removal effectiveness and, consequently, abrasivity. Even those formulations have abrasivity well within the safety limit regulatory agencies have established and, therefore, can be considered safe. Copyright © 2017 American Dental Association. Published by Elsevier Inc. All rights reserved.

Nevada State plan; final approval determination. Occupational Safety and Health Administration (OSHA), U.S. Department of Labor. Final State plan approval--Nevada.

Science.gov (United States)

2000-04-18

This document amends OSHA's regulations to reflect the Assistant Secretary's decision granting final approval to the Nevada State plan. As a result of this affirmative determination under section 18(e) of the Occupational Safety and Health Act of 1970, Federal OSHA's standards and enforcement authority no longer apply to occupational safety and health issues covered by the Nevada plan, and authority for Federal concurrent jurisdiction is relinquished. Federal enforcement jurisdiction is retained over any private sector maritime employment, private sector employers on Indian land, and any contractors or subcontractors on any Federal establishment where the land is exclusive Federal jurisdiction. Federal jurisdiction remains in effect with respect to Federal government employers and employees. Federal OSHA will also retain authority for coverage of the United States Postal Service (USPS), including USPS employees, contract employees, and contractor-operated facilities engaged in USPS mail operations.

Advanced fuels safety comparisons

International Nuclear Information System (INIS)

Grolmes, M.A.

1977-01-01

The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

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

International Nuclear Information System (INIS)

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

2007-01-01

The international nuclear community has expressed concern that some changes in existing plants could challenge safety margins while fulfilling all the regulatory requirements. In 1998, NEA published a report by the Committee on Nuclear Regulatory Activities on Future Nuclear Regulatory Challenges. The report recognized 'Safety margins during more exacting operating modes' as a technical issue with potential regulatory impact. Examples of plant changes that can cause such exacting operating modes include power up-rates, life extension or increased fuel burnup. In addition, the community recognized that the cumulative effects of simultaneous changes in a plant could be larger than the accumulation of the individual effects of each change. In response to these concerns, CSNI constituted the safety margins action plan (SMAP) task group with the following objectives: 'To agree on a framework for integrated assessments of the changes to the overall safety of the plant as a result of simultaneous changes in plant operation / condition; To develop a CSNI document which can be used by member countries to assess the effect of plant change on the overall safety of the plant; To share information and experience.' The two approaches to safety analysis, deterministic and probabilistic, use different methods and have been developed mostly independently of each other. This makes it difficult to assure consistency between them. As the trend to use information on risk (where the term risk means results of the PSA/PRA analysis) to support regulatory decisions is growing in many countries, it is necessary to develop a method of evaluating safety margin sufficiency that is applicable to both approaches and, whenever possible, integrated in a consistent way. Chapter 2 elaborates on the traditional view of safety margins and the means by which they are currently treated in deterministic analyses. This chapter also discusses the technical basis for safety limits as they are used today

Cooperation between the French and German safety authorities: a process of progress

International Nuclear Information System (INIS)

Lacoste, A.C.; Hennenhoefer, G.

1996-01-01

Considerable efforts are devoted in several countries worldwide for the designing of a new generation of electro-nuclear reactors with noticeable safety improvements with respect to the existing units. The EPR (European Pressurized water Reactor) project is the first example of a joint survey and collaboration effort between two countries, France and Germany, both characterized by a strong nuclear industry and competence. Franco-German relations have considerably expanded in recent years, enabling the French and German electricians, manufacturers (EDF, Siemens, Framatome..), and safety authorities (DSIN, IPSN, BMU, GRS..) to adopt common positions and decisions on fundamental topics about technical aspects of safety. This kind of collaboration and alignment was obviously only possible because two conditions were fulfilled: the discussion concerned future reactors and not those already built for which safety rules and provisions are not easily modified; and a common industrial project existed and had consequently to comply with jointly defined safety objectives. These two conditions are still fulfilled, so it should be possible for the work between French and German safety authorities to continue successfully, even in the next more delicate stage, involving definition of common positions on more detailed technical provisions. Finally, it is unthinkable that this important work undertaken by France and Germany could be carried out without frequent consultation of the safety authorities of other nuclear countries, notably in Europe. (J.S.). 1 fig

Project Guarantee 1985. Final repository for low- and intermediate-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

The safety barrier system for the type B repository for low- and intermediate-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). In the case of low- and intermediate-level waste the technical safety barrier system comprises: waste solidification matrix (cement, bitumen and resin), immobilisation of the waste packages in containers using liquid cement, concrete repository containers, backfilling of remaining vacant storage space with special concrete, concrete lining of the repository caverns, sealing of access tunnels on final closure of the repository. Natural geological safety barriers - host rock and overlying formations - have the following important functions. Because of its stability, the host rock in the repository zone protects the technical safety barrier system from destruction caused by climatic effects and erosion for a sufficient length of time. It also provides for low water flow and favourable chemistry (reducing conditions)

Final disposal of spent fuel in the Finnish bedrock. Scope and requirements for site-specific safety analysis

International Nuclear Information System (INIS)

1996-12-01

The report is a summary of the research conducted in the period 1993 to 1996 into safety of spent fuel final disposal. The principal goal of the research in this period, as set in 1993, was to develop a strategy for site-specific safety analysis. At the same time efforts were to be continued to gather data and validate the technical approach for the analysis. The work aimed at having the data needed for the analysis available at the end of year 1998. A safety assessment update, TILA-96, prepared by VTT Energy, is published as a separate report. The assessment is based on the TVO-92 safety analysis, but takes into account the knowledge acquired after 1992 on safety aspects of the disposal system and the data gathered from the site investigations made by TVO and from the beginning of 1996, by Posiva. Since the site investigations are still ongoing and much of the data gathered still pending interpretation, only limited amount of new site-specific information has been available for the present assessment. (172 refs.)

LOFT integral test system final safety analysis report

International Nuclear Information System (INIS)

1974-03-01

Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-15

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

78 FR 27419 - Final Safety Culture Policy Statement

Science.gov (United States)

2013-05-10

... organizational characteristics are present in a culture that promotes safety and environmental responsibility. A characteristic, in this case, is a pattern of thinking, feeling, and behaving that emphasizes safety...

Basic consideration on safety of facilities for final disposal of radioactive wastes, in particular for wastes stored in Abadia de Goias

International Nuclear Information System (INIS)

Xavier, A.M.; Mezrahi, A.; Heilbron Filho, P.F.L.

1991-01-01

The aim of this work is to contribute to the best understanding of aspects related to the safety criteria applied to repositories for radioactive wastes, in particular for wastes from the radiological accident occured in Goiania (Brazil) in September, 1987. (E.O.)

Conducting organizational safety reviews - requirements, methods and experience

International Nuclear Information System (INIS)

Reiman, T.; Oedewald, P.; Wahlstroem, B.; Rollenhagen, C.; Kahlbom, U.

2008-03-01

Organizational safety reviews are part of the safety management process of power plants. They are typically performed after major reorganizations, significant incidents or according to specified review programs. Organizational reviews can also be a part of a benchmarking between organizations that aims to improve work practices. Thus, they are important instruments in proactive safety management and safety culture. Most methods that have been used for organizational reviews are based more on practical considerations than a sound scientific theory of how various organizational or technical issues influence safety. Review practices and methods also vary considerably. The objective of this research is to promote understanding on approaches used in organizational safety reviews as well as to initiate discussion on criteria and methods of organizational assessment. The research identified a set of issues that need to be taken into account when planning and conducting organizational safety reviews. Examples of the issues are definition of appropriate criteria for evaluation, the expertise needed in the assessment and the organizational motivation for conducting the assessment. The study indicates that organizational safety assessments involve plenty of issues and situations where choices have to be made regarding what is considered valid information and a balance has to be struck between focus on various organizational phenomena. It is very important that these choices are based on a sound theoretical framework and that these choices can later be evaluated together with the assessment findings. The research concludes that at its best, the organizational safety reviews can be utilised as a source of information concerning the changing vulnerabilities and the actual safety performance of the organization. In order to do this, certain basic organizational phenomena and assessment issues have to be acknowledged and considered. The research concludes with recommendations on

Conducting organizational safety reviews - requirements, methods and experience

Energy Technology Data Exchange (ETDEWEB)

Reiman, T.; Oedewald, P.; Wahlstroem, B. [Technical Research Centre of Finland, VTT (Finland); Rollenhagen, C. [Royal Institute of Technology, KTH, (Sweden); Kahlbom, U. [RiskPilot (Sweden)

2008-03-15

Organizational safety reviews are part of the safety management process of power plants. They are typically performed after major reorganizations, significant incidents or according to specified review programs. Organizational reviews can also be a part of a benchmarking between organizations that aims to improve work practices. Thus, they are important instruments in proactive safety management and safety culture. Most methods that have been used for organizational reviews are based more on practical considerations than a sound scientific theory of how various organizational or technical issues influence safety. Review practices and methods also vary considerably. The objective of this research is to promote understanding on approaches used in organizational safety reviews as well as to initiate discussion on criteria and methods of organizational assessment. The research identified a set of issues that need to be taken into account when planning and conducting organizational safety reviews. Examples of the issues are definition of appropriate criteria for evaluation, the expertise needed in the assessment and the organizational motivation for conducting the assessment. The study indicates that organizational safety assessments involve plenty of issues and situations where choices have to be made regarding what is considered valid information and a balance has to be struck between focus on various organizational phenomena. It is very important that these choices are based on a sound theoretical framework and that these choices can later be evaluated together with the assessment findings. The research concludes that at its best, the organizational safety reviews can be utilised as a source of information concerning the changing vulnerabilities and the actual safety performance of the organization. In order to do this, certain basic organizational phenomena and assessment issues have to be acknowledged and considered. The research concludes with recommendations on

Fusion facility siting considerations

International Nuclear Information System (INIS)

Bussell, G.T.

1985-01-01

Inherent in the fusion program's transition from hydrogen devices to commercial power machines is a general increase in the size and scope of succeeding projects. This growth will lead to increased emphasis on safety, environmental impact, and the external effects of fusion in general, and of each new device in particular. A critically important consideration in this regard is site selection. The purpose of this paper is to examine major siting issues that may affect the economics, safety, and environmental impact of fusion

JET Tokamak, preparation of a safety case for tritium operations

Energy Technology Data Exchange (ETDEWEB)

Boyer, Helen, E-mail: helen.boyer@ccfe.ac.uk [CCFE, Culham Science Centre (United Kingdom); Plummer, David; Johnston, Jane [CCFE, Culham Science Centre (United Kingdom)

2016-11-01

Highlights: • A safety case incorporating technical and ITER related upgrades. • Hazard analysis reworked to include new modelling assessments. • Fitness for purpose assessment of safety controls. - Abstract: A new Safety Case is required to permit tritium operations on JET during the forthcoming DTE2 campaign. The outputs, benefits and lessons learned associated with the production of this Safety Case are presented. The changes that have occurred to the Safety Case methodology since the last JET tritium Safety Case are reviewed. Consideration is given to the effects of modifications, particularly ITER related changes, made to the JET and the impact these have on the hazard assessments as well as normal operations. Several specialized assessments, including recent MELCOR modelling, have been undertaken to support the production of this Safety Case and the impact of these assessments is outlined. Discussion of the preliminary actions being taken to progress implementation of this Safety Case is provided, highlighting new methods to improve the dissemination of the key Safety Case results to the plant operators. Finally, the work required to complete this Safety Case, before the next tritium campaign, is summarized.

Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

International Nuclear Information System (INIS)

Filbert, Wolfgang; Herold, Philipp

2015-01-01

The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.

Safety for fuel assembly handling in the nuclear ship Mutsu

International Nuclear Information System (INIS)

Ando, Yoshio

1978-01-01

The safety for fuel assembly handling in the nuclear ship Mutsu is deliberated by the committee of general inspection and repair technique examination for Mutsu. The result of deliberation for both cases of removing fuel assemblies and keeping them in the reactor is outlined. The specification of fuel assemblies, and the nuclides and designed radioactivity of fission products of fuel are described. The possibility of shielding repair work and general safety inspection keeping the fuel assemblies in the reactor, the safety consideration when the fuel assemblies are removed at a quay, in a dry dock and on the ocean, the safety of fuel transport in special casks and fuel storage are explained. It is concluded finally that the safety of shielding repair work and general inspection work is secured when the fuel assemblies are kept in the reactor and also when the fuel assemblies are removed from the reactor by cautious working. (Nakai, Y.)

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.)

Human factors in safety assessment. Safety culture assessment

International Nuclear Information System (INIS)

Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

1996-01-01

This paper analyses the present conditions and problems in enterprises safety assessment, and introduces the characteristics and effects of safety culture. The authors think that safety culture must be used as a 'soul' to form the pattern of modern safety management. Furthermore, they propose that the human safety and synthetic safety management assessment in a system should be changed into safety culture assessment. Finally, the assessment indicators are discussed

CTLA-4 blockade with ipilimumab: biology, safety, efficacy, and future considerations

International Nuclear Information System (INIS)

Camacho, Luis H

2015-01-01

Melanoma remains a critical public health problem worldwide. Patients with stage IV disease have very poor prognosis and their 1-year survival rate is only 25%. Until recently, systemic treatments with a positive impact on overall survival (OS) had remained elusive. In recent years, the United States Food and Drug Administration (FDA) – approved several novel agents targeting the RAS/RAF/MEK/ERK pathway (vemurafenib, dabrafenib, and trametinib) – critical in cell division and proliferation of melanoma, and an immune checkpoint inhibitor (ipilimumab) directed against the cytotoxic T lymphocyte Antigen - (CTLA-4). Moreover, recent reports of clinical trials studying other immune checkpoint modulating agents will most likely result in their FDA approval within the next months. This review focuses on ipilimumab, its safety and efficacy, and future considerations. Ipilimumab has demonstrated a positive OS impact after a several-year follow-up. It is also recognized that due to its mechanism of action, the response patterns to ipilimumab can differ from those observed in patients following treatment with conventional cytotoxic agents and even the most recently approved BRAF inhibitors. Most patients (84.8%) experience drug-related adverse events (AEs) of any grade; most of these are mild to moderate and immune mediated. However, a minority of patients may also experience severe and life-threatening AEs. In clinical studies, AEs were managed according to guidelines that emphasized close clinical monitoring and early use of corticosteroids when appropriate. Preliminary results have taught us the potential greater toxicity when in combination with vemurafenib, and the greater antitumor efficacy when combined with nivolumab, a monoclonal antibody directed against programmed death receptor-1 (PD-1), another immune checkpoint inhibitor. Future challenges include the optimization of dosing and toxicities when used as a single agent, and studying the safety and efficacy of

Community Road Safety Initiatives for the Minerals Industry

Directory of Open Access Journals (Sweden)

Tim Horberry

2013-12-01

Full Text Available Major companies in the minerals industry are increasingly recognizing that their operations have an impact in the wider community. Regarding transportation issues, this impact extends beyond purely the safety of company vehicle fleets to consideration of Community Road Safety (CRS concerns, which address the driving, walking, and riding practices of community members in a locale with increased heavy vehicle traffic. Our assessment here of national and international trends in approaches to road safety awareness and associated road safety strategies is meant to inform companies in the minerals industry of developments that can influence the design of their road safety initiatives. The review begins by considering the overall road safety context and the dominant “safe systems” framework employed internationally. Thereafter, it considers what is typically included in CRS initiatives for the minerals industry. Three case studies are then presented to highlight approaches that feature exemplary collaboration, design, implementation, or impact. Thereafter, we analyze lessons learnt by key researchers and practitioners in the CRS field. Finally, we conclude that best CRS practices for the minerals industry rely on eleven factors, including for example collaboration with local entities and stepwise implementation.

Full-Length High-Temperature Severe Fuel Damage Test No. 5: Final safety analysis

International Nuclear Information System (INIS)

Lanning, D.D.; Lombardo, N.J.; Panisko, F.E.

1993-09-01

This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the Full-Length High Temperature Experiment-5 (FLHT-5) to be conducted in the L-24 position of the National Research Universal (NRU) Reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test is sponsored by an international group organized by the US Nuclear Regulatory Commission. The test is designed and conducted by staff from Pacific Northwest Laboratory with CRNL staff support. The test will study the consequences of loss-of-coolant and the progression of severe fuel damage

Bipolar disorder and complementary medicine: current evidence, safety issues, and clinical considerations.

Science.gov (United States)

Sarris, Jerome; Lake, James; Hoenders, Rogier

2011-10-01

Bipolar disorder (BD) is a debilitating syndrome that is often undiagnosed and undertreated. Population surveys show that persons with BD often self-medicate with complementary and alternative medicine (CAM) or integrative therapies in spite of limited research evidence supporting their use. To date, no review has focused specifically on nonconventional treatments of BD. The study objectives were to present a review of nonconventional (complementary and integrative) interventions examined in clinical trials on BD, and to offer provisional guidelines for the judicious integrative use of CAM in the management of BD. PubMed, CINAHL,(®) Web of Science, and Cochrane Library databases were searched for human clinical trials in English during mid-2010 using Bipolar Disorder and CAM therapy and CAM medicine search terms. Effect sizes (Cohen's d) were also calculated where data were available. Several positive high-quality studies on nutrients in combination with conventional mood stabilizers and antipsychotic medications in BD depression were identified, while branched-chain amino acids and magnesium were effective (small studies) in attenuating mania in BD. In the treatment of bipolar depression, evidence was mixed regarding omega-3, while isolated studies provide provisional support for a multinutrient formula, n-acetylcysteine, and l-tryptophan. In one study, acupuncture was found to have favorable but nonsignificant effects on mania and depression outcomes. Current evidence supports the integrative treatment of BD using combinations of mood stabilizers and select nutrients. Other CAM or integrative modalities used to treat BD have not been adequately explored to date; however, some early findings are promising. Select CAM and integrative interventions add to established conventional treatment of BD and may be considered when formulating a treatment plan. It is hoped that the safety issues and clinical considerations addressed in this article may encourage the practice

Final Hazard Classification and Auditable Safety Analysis for the 105-F Building Interim Safe Storage Project

International Nuclear Information System (INIS)

Rodovsky, T.J.; Bond, S.L.

1998-07-01

The auditable safety analysis (ASA) documents the authorization basis for the partial decommissioning and facility modifications to place the 105-F Building into interim safe storage (ISS). Placement into the ISS is consistent with the preferred alternative identified in the Record of Decision (58 FR). Modifications will reduce the potential for release and worker exposure to hazardous and radioactive materials, as well as lower surveillance and maintenance (S ampersand M) costs. This analysis includes the following: A description of the activities to be performed in the course of the 105-F Building ISS Project. An assessment of the inventory of radioactive and other hazardous materials within the 105-F Building. Identification of the hazards associated with the activities of the 105-F Building ISS Project. Identification of internally and externally initiated accident scenarios with the potential to produce significant local or offsite consequences during the 105-F Building ISS Project. Bounding evaluation of the consequences of the potentially significant accident scenarios. Hazard classification based on the bounding consequence evaluation. Associated safety function and controls, including commitments. Radiological and other employee safety and health considerations

Alcator C-MOD final safety analysis

International Nuclear Information System (INIS)

Fiore, C.L.

1989-06-01

This document is designed to address the safety issues involved with the Alcator C-Mod project. This report will begin with a brief description of the experimental objectives which will be followed by information concerning the site. The Alcator C-Mod experiment is a pulsed fusion experiment in which a plasma formed from small amounts of hydrogen or deuterium gas is confined in a magnetic field for short periods (∼1 s). No radioactive fuels or fissile materials are used in the device, so that no criticality hazard exists and no credible nuclear accident can occur. During deuterium operation, the production of a small number of neutrons from a short pulse could result in a small amount of short- and intermediate-lived radioactive isotopes being produced inside the experimental cell. This report will demonstrate that this does not pose an additional hazard to the general population. The health and safety hazards resulting from Alcator C-Mod occur to the workers on the experiment, each of which is described in its own chapter with the steps taken to minimize the risk to employees. These hazards include fire, chemicals and cryogenics, air quality, electrical, electromagnetic radiation, ionizing radiation, and mechanical and natural phenomena. None of these hazards is unique to the facility, and methods of protection from them are well defined and are discussed in the chapter which describes each hazard. The quality assurance program, critical to ensuring the safety aspects of the program, will also be described

Multi-objective demand side scheduling considering the operational safety of appliances

International Nuclear Information System (INIS)

Du, Y.F.; Jiang, L.; Li, Y.Z.; Counsell, J.; Smith, J.S.

2016-01-01

Highlights: • Operational safety of appliances is introduced in multi-objective scheduling. • Relationships between operational safety and other objectives are investigated. • Adopted Pareto approach is compared with Weigh and Constraint approaches. • Decision making of Pareto approach is proposed for final appliances’ scheduling. - Abstract: The safe operation of appliances is of great concern to users. The safety risk increases when the appliances are in operation during periods when users are not at home or when they are asleep. In this paper, multi-objective demand side scheduling is investigated with consideration to the appliances’ operational safety together with the electricity cost and the operational delay. The formulation of appliances’ operational safety is proposed based on users’ at-home status and awake status. Then the relationships between the operational safety and the other two objectives are investigated through the approach of finding the Pareto-optimal front. Moreover, this approach is compared with the Weigh and Constraint approaches. As the Pareto-optimal front consists of a set of optimal solutions, this paper proposes a method to make the final scheduling decision based on the relationships among the multiple objectives. Simulation results demonstrate that the operational safety is improved with the sacrifice of the electricity cost and the operational delay, and that the approach of finding the Pareto-optimal front is effective in presenting comprehensive optimal solutions of the multi-objective demand side scheduling.

Two-phase flow in a saliniferous final repository using the example of ERAM. Final report; Zweiphasenfluss in einem salinaren Endlager am Beispiel des ERAM. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Kock, Ingo; Frieling, Gerd; Navarro, Martin

2016-10-15

In the frame of the research project ZIESEL the GRS enhanced the state of science and technology for the realization and evaluation of long-term safety cases for the final deposition of radioactive wastes. The superior aim was the improved understanding of two-phase flow processes in a complex final repository system. The consideration of two-phase processes in modeling of final repository systems induces processes and effects that significantly affect the transport behavior of fluid and radionuclides. Two-phase processes include not only capillary pressures and relative permeabilities but also a basic competition of phases with respect to pore volume for storage and transport and density-driven vertical separation of phases. Basically seals have been shown to be essential for the system behavior because of their influence of the gas pressure dependent control function. The system behavior is also influences by the model geometry.

Integrated plant safety assessment. Systematic evaluation program, Big Rock Point Plant (Docket No. 50-155). Final report

International Nuclear Information System (INIS)

1984-05-01

The Systematic Evaluation Program was initiated in February 1977 by the U.S. Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety when the supplement to the Final Integrated Plant Safety Assessment Report has been issued. This report documents the review of the Big Rock Point Plant, which is one of ten plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addressed. It also addresses a majority of the pending licensing actions for Big Rock Point, which include TMI Action Plan requirements and implementation criteria for resolved generic issues. Equipment and procedural changes have been identified as a result of the review

Systems Analysis of NASA Aviation Safety Program: Final Report

Science.gov (United States)

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

2013-01-01

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

49 CFR 190.323 - Consideration of comments received.

Science.gov (United States)

2010-10-01

... 49 Transportation 3 2010-10-01 2010-10-01 false Consideration of comments received. 190.323 Section 190.323 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND... SAFETY PROGRAMS AND RULEMAKING PROCEDURES Procedures for Adoption of Rules § 190.323 Consideration of...

Technical basis for the ITER final design report, cost review and safety analysis (FDR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-12-01

The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement. Refs, figs, tabs

Technical basis for the ITER final design report, cost review and safety analysis (FDR)

International Nuclear Information System (INIS)

1998-01-01

The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

International Nuclear Information System (INIS)

Kilger, R.; Gmal, B.; Moser, E.F.

2008-01-01

Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

49 CFR 389.23 - Consideration of comments received.

Science.gov (United States)

2010-10-01

... 49 Transportation 5 2010-10-01 2010-10-01 false Consideration of comments received. 389.23 Section 389.23 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL MOTOR CARRIER... PROCEDURES-FEDERAL MOTOR CARRIER SAFETY REGULATIONS Procedures for Adoption of Rules § 389.23 Consideration...

Final hazard classification and auditable safety analysis for the N basin segment

International Nuclear Information System (INIS)

Kloster, G.; Smith, R.I.; Larson, A.R.; Duncan, G.M.

1996-12-01

The purpose of this report is to provide the following: To serve as the auditable safety analysis (ASA) for the N Basin Segment, including both the quiescent state and planned intrusive activities. The ASA is developed through the realistic evaluation of potential hazards that envelope the threat to personnel. The ASA also includes the specification of the programmatic, baseline, and activity- specific controls that are necessary for the protection of workers. To determine and document the final hazard classification (FHC) for the N Basin Segment. The FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel. The FHC also includes the specification of the special controls that are necessary to remain within the envelope of those accident analyses

Considerations concerning the reliability of reactor safety equipment; Considerations sur la fiabilite des ensembles de securite de reacteurs

Energy Technology Data Exchange (ETDEWEB)

Furet, J; Guyot, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [French] On rappelle les circonstances qui favorisent au C.E.A. la collecte d'une information valable des resultats de la maintenance. L'importance des donnees a traiter a rendu necessaire l'utilisation d'une calculatrice poux l'analyse automatique des resultats recueillis. On se limitera ici aux aspects particuliers de la fiabilite du point de vue de l'electronique pour le controle et la commande de reacteurs nucleaires: pannes sures et pannes non sures; probabilite de survie dans le cas de la securite des reacteurs; facteur de disponibilite. Les schemas de principe des ensembles de securite definis pour deux types de reacteurs (reacteur de puissance et reacteur experimental de faible puissance) sont indiques. On

Final hazard classification and auditable safety analysis for the 105-C Reactor Interim Safe Storage Project

International Nuclear Information System (INIS)

Rodovsky, T.J.; Larson, A.R.; Dexheimer, D.

1996-12-01

This document summarizes the inventories of radioactive and hazardous materials present in the 105-C Reactor Facility and the operations associated with the Interim Safe Storage Project which includes decontamination and demolition and interim safe storage of the remaining facility. This document also establishes a final hazard classification and verifies that appropriate and adequate safety functions and controls are in place to reduce or mitigate the risk associated with those operations

Safety First

Science.gov (United States)

Taft, Darryl

2011-01-01

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

Safety considerations for the use of nuclear power in space

International Nuclear Information System (INIS)

Sewell, D.C.

1985-01-01

A little over twenty years ago Norris Bradbury, then Director of the Los Alamos Scientific Laboratory, gave a luncheon speech at the American Nuclear Society Meeting on Aerospace Nuclear Safety here in Albuquerque, New Mexico. His subject was Safety in Science. His opening statement is well worth recalling. He said, ''...science - by which I mean both science and technology - has historically generated new hazards and, equally historically, invented new safeties for mankind. It has produced a safer way of life, a lesser dependence on man's physical frailities.'' He went on to say, ''Wherever science has made an advance it has developed a new hazard, but in general the new hazards have been less overall than the hazards made obsolete by the new development.'' I think that these are excellent thoughts to keep in mind as we embark on a program for increased use of nuclear power in space. That does not mean that the safe operation of new nuclear space power systems will come automatically. It will not. We must work at it continually to make these new systems meet acceptable safety standards

Laser safety considerations for a mobile laser program

Science.gov (United States)

Flor, Mary

1997-05-01

An increased demand for advanced laser technology, especially in the area of cutaneous and cosmetic procedures has prompted physicians to use mobile laser services. Utilization of a mobile laser service allows physicians to provide the latest treatments for their patients while minimizing overhead costs. The high capital expense of laser systems is often beyond the financial means of individual clinicians, group practices, free-standing clinics and smaller community hospitals. Historically rapid technology turnover with laser technology places additional risk which is unacceptable to many institutions. In addition, health care reform is mandating consolidation of equipment within health care groups to keep costs at a minimum. In 1994, Abbott Northwestern Hospital organized an in-house mobile laser technology service which employs a group of experienced laser specialists to deliver and support laser treatments for hospital outreach and other regional physicians and health care facilities. Many of the hospital's internal safety standards and policies are applicable to the mobile environment. A significant challenge is client compliance because of the delicate balance of managing risk while avoiding being viewed as a regulator. The clinics and hospitals are assessed prior to service to assure minimum laser safety standards for both the patient and the staff. A major component in assessing new sites is to inform them of applicable regulatory standards and their obligations to assure optimum laser safety. In service training is provided and hospital and procedures are freely shared to assist the client in establishing a safe laser environment. Physician and nursing preceptor programs are also made available.

SAFETY CONSIDERATIONS WITH BLOOD FLOW RESTRICTED RESISTANCE TRAINING

Directory of Open Access Journals (Sweden)

Alan Kacin

2015-11-01

Full Text Available Blood flow restricted resistance (BFRR training with pneumatic tourniquet has been suggested as an alternative for conventional weight training due to the proven benefits for muscle strength and hypertrophy using relatively low resistance, hence reducing the mechanical stress across a joint. As such, it has become an important part of rehabilitation programs used in either injured or operated athletes. Despite a general consensus on effectiveness of BFRR training for muscle conditioning, there are several uncertainties regarding the interplay of various extrinsic and intrinsic factors on its safety and efficiency, which are being reviewed from a clinical perspective. Among extrinsic factors tourniquet cuff pressure, size and shape have been identified as key for safety and efficiency. Among intrinsic factors, limb anthropometrics, patient history and presence of cardiac, vascular, metabolic or peripheral neurologic conditions have been recognized as most important. Though there are a few potential safety concerns connected to BFRR training, the following have been identified as the most probable and health-hazardous: (a mechanical injury to the skin, muscle, and peripheral nerves, (b venous thrombosis due to vascular damage and disturbed hemodynamics and (c augmented arterial blood pressure responses due to combined high body exertion and increased peripheral vascular resistance. Based on reviewed literature and authors’ personal experience with the use of BFRR training in injured athletes, some guidelines for its safe application are outlined. Also, a comprehensive risk assessment tool for screening of subjects prior to their inclusion in a BFRR training program is being introduced.

Fast reactor safety program. Progress report, January-March 1980

International Nuclear Information System (INIS)

1980-05-01

The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development

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

Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

International Nuclear Information System (INIS)

1983-04-01

The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

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

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.)

Nuclear safety considerations with emphasis on instrumentation and control systems

International Nuclear Information System (INIS)

Beare, J.W.

1978-01-01

The conceptual model of a nuclear power plant in Canada is that it consists basically of two kinds of systems. The first kind is the process systems, that is, those structures and components associated with the production of nuclear energy and its conversion to other forms of energy. The second kind is the special safety systems, whose purpose it is to protect the public in the event of a serious failure in the process systems which might otherwise lead to unacceptable radiological consequences. Quantitative limits are set on the unavailability of the special safety systems. These limits are low enough to be consistent with low overall risk and yet can be demonstrated by test during operation of the plant. Low unavailability is an important but not the only condition required for low unrealiability for the special safety systems. The special safety systems minimize the chance of a cross-linked failure particularly under the conditions experienced as a result of the more severe types of postulated serious process failures. Nuclear power plants must also withstand, without a major hazard to the public, certain rare events associated with natural phenomena or man-made activities off-site and also certain in-plant events such as fire or break-up of a turbine-generator which might have a cross-linking effect on process and safety systems. In the latest designs, Canadian nuclear power plants have emergency systems to deal with such events. The emergency systems have an enhanced degree of physical and functional separation from other plant systems. (author)

Fast Flux Test Facility final safety analysis report. Amendment 72

Energy Technology Data Exchange (ETDEWEB)

Gantt, D. A.

1992-08-01

This document provides the Final Safety Analysis Report (FSAR) Amendment 72 for incorporation into the Fast Flux Test Facility (FFTF) FSAR set. This amendment change incorporates Engineering Change Notices issued subsequent to Amendment 71 and approved for incorporation before June 24, 1992. These include changes in: Chapter 2, Site Characteristics; Chapter 3, Design Criteria Structures, Equipment, and Systems; Chapter 5B, Reactor Coolant System; Chapter 7, Instrumentation and Control Systems; Chapter 8, Electrical Systems - The description of the Class 1E, 125 Vdc systems is updated for the higher capacity of the newly installed, replacement batteries; Chapter 9, Auxiliary Systems - The description of the inert cell NASA systems is corrected to list the correct number of spare sample points; Chapter 11, Reactor Refueling System; Chapter 12, Radiation Protection and Waste Management; Chapter 13, Conduct of Operations; Chapter 16, Quality Assurance; Chapter 17, Technical Specifications; Chapter 19, FFTF Fire Specifications for Fire Detection, Alarm, and Protection Systems; Chapter 20, FFTF Criticality Specifications; and Appendix B, Primary Piping Integrity Evaluation.

Nuclear emergency preparedness. Final report of the Nordic Nuclear Safety Research Project BOK-1

DEFF Research Database (Denmark)

Lauritzen, B.

2002-01-01

Final report of the Nordic Nuclear Safety Research project BOK-1. The BOK-1 project, “Nuclear Emergency Preparedness”, was carried out in 1998-2001 with participants from the Nordic and Baltic Sea regions. The project consists of six sub-projects:Laboratory measurements and quality assurance (BOK-1.......1); Mobile measurements and measurement strategies (BOK-1.2); Field measurements and data assimilation (BOK-1.3); Countermeasures in agriculture and forestry (BOK-1.4); Emergency monitoring in theNordic and Baltic Sea countries (BOK-1.5); and Nuclear exercises (BOK-1.6). For each sub-project, the project...

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

Energy Technology Data Exchange (ETDEWEB)

ULLAH, M K

2001-02-26

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

Fire Safety Consideration in the Pre-conceptual Design State of Pyro-Facillity

Energy Technology Data Exchange (ETDEWEB)

Jeon, Hong Rae; Seo, Seok Jun; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The government, in order to solve this problem, has organized a public engagement committee and is searching for a solution. To use sustainable nuclear energy, our country is also pursuing research and development of fast breeder reactor and pyroprocessing technology in accordance with the international movement of spent fuel recycling and efforts towards nuclear non-proliferation which is centered on the development and demonstration of recycling spent fuel and fast breeder reactors. Pyro-facility has different features with nuclear power plant. In the pyroprocess, chemical and electrochemical separation were took place in the hot cells and material at risk (MAR) is distributed in many working areas. In this paper, we conducted the fire modeling of hot cells to see the stability of pyrophoric materials which is considered as one of the potential hazardous materials in the main process cell. Based on modeling results, consideration of fire safety pyrofacility will be discussed. We performed preliminary hazard analysis for pyrofacility and summarized potential fire hazard. Pyrophoric material fire is the dominant hazard in the main process hot cell and fire modeling of cable tray in the cell was analyzed to see the stability of pyrophoric materials. Analysis results clearly shows that pyrophoric materials are prone to be affected.

National Safety Council Final Report

International Nuclear Information System (INIS)

Norris, Karen; Shannon, Tom

2005-01-01

In December 1995, the National Safety Council (NSC) entered into Cooperative Agreement No.DE-FC02-96EW 12729 with the US Department of Energy (DOE) to work together over the next few years on safety and health initiatives surrounding the management of radioactive materials. As a result, three publications, including print and non-print deliverables, were developed and distributed: (1) Series of Backgrounders, Web Services for WIPP; (2) A Guide to Foreign Research Reactor Spent Fuel; and (3) A Guide to the US Department of Energy's Low-Level Radioactive Waste. DOE and its predecessor agencies have maintained a record of safe transportation of radioactive materials for more than 50 years. Thousands of shipments involving three million packages of radioactive materials are shipped each year in the United States. Historically, DOE shipments constitute less than one percent of the total radioactive material shipments; however, they comprise a significant portion (approaching 75 percent) of the curies, or amounts of radioactivity shipped annually. DOE operations and field offices are responsible for detailed planning and for ensuring full regulatory compliance for their shipments. Packaging is designed to protect workers and limit the risk to the public during transportation. DOE headquarters and program offices provide policy direction and oversight for packaging and transportation activities for their respective offices. The publications NSC produced under the agreement also included primary points of contact for external audiences, including the press, the public, and stakeholders who would not have access to DOE regulations, manuals, and practices

Identification of important ''PIUS'' design considerations and accident sequences using qualitative plant assessment techniques

International Nuclear Information System (INIS)

Higgins, J.; Fullwood, R.; Kroeger, P.; Youngblood, R.

1992-01-01

The PIUS (Process Inherent Ultimate Safety) reactor is an advanced design nuclear power plant that uses passive safety features and basic physical processes to address safety concerns. Brookhaven National Laboratory (BNL) performed a detailed study of the PIUS design for the NRC using primarily qualitative engineering analysis techniques. Some quantitative methods were also employed. There are three key initial areas of analysis: FMECA, HAZOP, and deterministic analyses, which are described herein. Once these three analysis methods were completed, the important findings from each of the methods were assembled into thePIUS Interim Table (PIT). This table thus contains a first cut sort of the important design considerations and features of the PIUS reactor. The table also identifies some potential initiating events and systems used for mitigating these initiators. The next stage of the analysis was the construction of event trees for each of the identified initiators. The most significant sequences were then determined qualitatively, using, some quantitative input. Finally, overall insights on the PIUS design developed from the PIT and from the event tree analysis were developed and presented

Safety and emergency preparedness considerations for geotechnical field operations

Energy Technology Data Exchange (ETDEWEB)

Wemple, R.P.

1989-04-01

The GEO Energy Technology Department at Sandia National Laboratories is involved in several remote-site drilling and/or experimental operations each year. In 1987, the Geothermal Research Division of the Department developed a general set of Safe Operating Procedures (SOPs) that could be applied to a variety of projects. This general set is supplemented by site-specific SOPs as needed. Effective field operations require: integration of safety and emergency preparedness planning with overall project planning, training of field personnel and inventorying of local emergency support resources, and, developing a clear line of responsibility and authority to enforce the safety requirements. Copies of SOPs used in recent operations are included as examples of working documents for the reader.

Strategy for public understanding and participation in nuclear safety regulation

Energy Technology Data Exchange (ETDEWEB)

Kim, H. J.; Chung, Yun Hyung

2004-02-15

The objective of this study is to help the general public and local residents to better understand and trust nuclear safety regulation. In order to obtain public confidence in nuclear safety regulation, the emotion and demand of public should be first understood and the change in an attitude to meet the present circumstances actively is requisite. Hence it is intended that a genuine communication shall be newly arranged and accomplished on the basis of mutual understanding. To achieve this, a series of public opinion poll have performed periodically and symposium for the public acceptance is held in order to frame a policy based on the understanding of nuclear safety and regulation of the general public and local residents. Besides nuclear safety indicators including safety sentiment indicators are being developed as a means to understand the safety of operating nuclear power plants from the viewpoint of the general public, a plan for the harmonious communication of nuclear safety information is established, and handbooks of nuclear terminologies and report-writing are under development in part. Finally plans for convergence of the public opinions and a wide public involvement in nuclear safety regulation are formulated and their applicability as organization and administration program is now under consideration.

Fusion reactor safety studies, FY 1977

International Nuclear Information System (INIS)

Darby, J.B. Jr.

1978-04-01

This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

Safety considerations in providing allergen immunotherapy in the office.

Science.gov (United States)

Mattos, Jose L; Lee, Stella

2016-06-01

This review highlights the risks of allergy immunotherapy, methods to improve the quality and safety of allergy treatment, the current status of allergy quality metrics, and the future of quality measurement. In the current healthcare environment, the emphasis on outcomes measurement is increasing, and providers must be better equipped in the development, measurement, and reporting of safety and quality measures. Immunotherapy offers the only potential cure for allergic disease and asthma. Although well tolerated and effective, immunotherapy can be associated with serious consequence, including anaphylaxis and death. Many predisposing factors and errors that lead to serious systemic reactions are preventable, and the evaluation and implementation of quality measures are crucial to developing a safe immunotherapy practice. Although quality metrics for immunotherapy are in their infancy, they will become increasingly sophisticated, and providers will face increased pressure to deliver safe, high-quality, patient-centered, evidence-based, and efficient allergy care. The establishment of safety in the allergy office involves recognition of potential risk factors for anaphylaxis, the development and measurement of quality metrics, and changing systems-wide practices if needed. Quality improvement is a continuous process, and although national allergy-specific quality metrics do not yet exist, they are in development.

Nuclear Electric flask design and safety case development during the last ten years

International Nuclear Information System (INIS)

Dougall, I.; Jones, D.K.

1994-01-01

The scope of Nuclear Electric's (NE) requirements for irradiated fuel transport and the relevant safety and regulatory compliance standards are outlined. To illustrate NE's approach to demonstrating package design compliance the basis of the safety case for unbottled Magnox fuel transport in the Mk M2 Magnox flask is described. The considerations which lay behind the development of the Mk A2 AGR Flask are indicated and a description is given of the main design features of this flask. Finally there is a summary of progress in obtaining Type B(M) approvals, based on the 1985 IAEA Regulations, for the requisite range of flask contents. (author)

Waste compatibility safety issues and final results for tank 241-T-110 push mode samples

International Nuclear Information System (INIS)

Nuzum, J.L.

1997-01-01

This document is the final laboratory report for Tank 241-T-110. Push mode core segments were removed from risers 2 and 6 between January 29, 1997, and February 7, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-T-110 Push Mode Core Sampling and analysis Plan (TSAP) and Safety Screening Data Quality Objective (DQO). None of the subsamples submitted for total alpha activity (AT) or differential scanning calorimetry (DSC) analyses exceeded the notification limits stated in DQO

Technical considerations associated with spent fuel acceptance. Final report

International Nuclear Information System (INIS)

Supko, E.M.

1996-06-01

This study was initiated by the Electric Power Research Institute (EPRI) to identify technical considerations associated with spent fuel acceptance and implementation of a waste management system that includes the use of transportable storage systems, and to serve as an opening dialogue among Standard Contract Holders and the department of Energy's Office of Civilian Radioactive Waste management (OCRWM) prior to the development of waste acceptance criteria or issuance of a Notice of Proposed Rulemaking by OCRWM to amend the Standard Contract. The original purpose of the Notice of Proposed Rulemaking was to address changes to the Standard Contract to implement a multi-purpose canister based system and to address other issues that were not adequately addressed in the standard contract. Even if DOE does not develop a multi-purpose canister based system for waste acceptance, it will still be necessary to develop waste acceptance criteria in order to accept spent fuel in transportable storage systems that are being deployed for at-reactor storage. In this study, technical issues associated with spent fuel acceptance will be defined and potential options and alternatives for resolution of technical considerations will be explored

NRC safety research in support of regulation

International Nuclear Information System (INIS)

1994-06-01

This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1993. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with sound technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission's rules, regulatory guides, or other guidance

Safety first. Yes, but which safety?; Primat der Sicherheit. Ja, aber welche Sicherheit ist gemeint?

Energy Technology Data Exchange (ETDEWEB)

Roehlig, Klaus-Juergen [Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany). Inst. fuer Endlagerforschung; Eckhardt, Anne [risicare GmbH, Zollikerberg (Switzerland)

2017-09-01

The site selection law in Germany and the final report of the final repository commission state the central objective to find a repository site that will guarantee safety for the next million of years. Decision makers, concerned and interested people have obviously different opinions and acceptance criteria with respect to the tools for the demonstration of safety (safety case). Possible solutions for a broad acceptance of safety definitions are discussed.

Safety-evaluation report related to the final design of the Standard Nuclear Steam Supply Reference System - CESSAR System 80. Docket No. STN 50-470

International Nuclear Information System (INIS)

1983-03-01

Supplement No. 1 to the Safety Evaluation Report for the application filed by Combustion Engineering, Inc. for a Final Design Approval for the Combustion Engineering Standard Safety Analysis Report (STN 50-470) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation by providing: (1) the evaluation of additional information submitted by the applicant since the Safety Evaluation Report was issued, (2) the evaluation of the matters the staff had under review when the Safety Evaluation Report was issued, and (3) the response to comments made by the Advisory Committee on Reactor Safeguards

Safety technical considerations on the 2012 periodic safety verification of the Beznau nuclear power plant; Sicherheitstechnische Stellungnahme zur Periodischen Sicherheitsüberprüfung 2012 des Kernkraftwerks Beznau

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-12-15

According to nuclear legislation, the owner of an operational license for a nuclear power plant has to provide a periodic safety verification (PSU) every 10 years. The 'North Eastern Power Plants' company (NOK), today AXPO Power AG already performed such a PSU for the Beznau-2 nuclear reactor block (KKB2) in 2002. The Beznau-1 nuclear reactor block (KKB1) received its definitive operational license in October 1970, after test operation during 7 months. After the license for test operation received on July 16{sup th}, 1971, the operational license of KKB2 was renewed several times, each time for a certain period of validity. In 1991, NOK requested a definitive operational license for KKB2, but in 1994 the Swiss Federal Council lengthened the license for only 10 years. Moreover, it laid down that NOK has to periodically report on the safety of the facility. With its letter of August 23{sup rd}, 1998, the Federal Office of Energy defined the documents to be produced for the PSU. The extent of the PSU was defined in such a way that many documents concern the whole power plant, i.e. both nuclear reactor blocks. On December 3{sup rd}, 2004, the Swiss Federal Council granted KKB2 an operational license of limited validity. The present report reviews the 2012 PSU, which covers the time interval from January 1{sup st}, 2002, to December 31{sup st}, 2011, from the point of view of safety. It contains documents for the evaluation of both reactor blocks at KKB. The Beznau interim storage pool was also taken into consideration; it is situated on the KKB site, but, according to a decision of the Swiss Federal Council of May 23{sup rd}, 1991, it has an independent operational license. The evaluation of ageing surveillance takes the whole operational period of the facility into account, i.e. the ageing mechanisms acting as from the beginning of the operation. Moreover, important developments that occurred after the surveillance time interval have been taken into account

Planning and architectural safety considerations in designing nuclear power plants

International Nuclear Information System (INIS)

Konsowa, Ahmed A.

2009-01-01

To achieve optimum safety and to avoid possible hazards in nuclear power plants, considering architectural design fundamentals and all operating precautions is mandatory. There are some planning and architectural precautions should be considered to achieve a high quality design and construction of nuclear power plant with optimum safety. This paper highlights predicted hazards like fire, terrorism, aircraft crash attacks, adversaries, intruders, and earthquakes, proposing protective actions against these hazards that vary from preventing danger to evacuating and sheltering people in-place. For instance; using safeguards program to protect against sabotage, theft, and diversion. Also, site and building well design focusing on escape pathways, emergency exits, and evacuation zones, and the safety procedures such as; evacuation exercises and sheltering processes according to different emergency classifications. In addition, this paper mentions some important codes and regulations that control nuclear power plants design, and assessment methods that evaluate probable risks. (author)

International nuclear safety

International Nuclear Information System (INIS)

Wolff, P.H.W.

1978-01-01

The background to the development of internationally agreed safety principles and practices is discussed. The activities of the IAEA and the scope, structure, and organisation of its programme of Reactor Safety Codes and Guides are described. Attention is drawn to certain areas needing further considerations. (UK)

Decommissioning considerations at a time of nuclear renaissance

International Nuclear Information System (INIS)

Devgun, Jas S.

2007-01-01

At a time of renaissance in the nuclear power industry, when it is estimated that anywhere between 60 to 130 new power reactors may be built worldwide over the next 15 years, why should we focus on decommissioning? Yet it is precisely the time to examine what decommissioning considerations should be taken into account as the industry proceeds with developing final designs for new reactors and the construction on the new build begins. One of the lessons learned from decommissioning of existing reactors has been that decommissioning was not given much thought when these reactors were designed three or four decades ago. Even though decommissioning may be sixty years down the road from the time they go on line, eventually all reactors will be decommissioned. It is only prudent that new designs be optimized for eventual decommissioning, along with the other major considerations. The overall objective in this regard is that when the time comes for decommissioning, it can be completed in shorter time frames, with minimum generation of radioactive waste, and with better radiological safety. This will ensure that the tail end costs of the power reactors are manageable and that the public confidence in the nuclear power is sustained through the renaissance and beyond. (author)

Safety aspects with regard to plutonium vitrification techniques

International Nuclear Information System (INIS)

Gray, L.W.; Kan, T.

1995-01-01

Substantial inventories of excess plutonium are expected to result from dismantling US and Russian nuclear weapons. Disposition of this material should be a high priority in both countries. Various disposition options are under consideration. One option is to vitrify the plutonium with the addition of 137 Cs or high-level waste to act as a deterrent to proliferation. The primary safety problem associated with vitrification of plutonium is to avoid criticality in form fabrication and in the final repository over geologic time. Recovery should be as difficult (costly) as the recovery of plutonium from spent fuel

Collection and accumulation of seismic safety research findings, and considerations for information dissemination

International Nuclear Information System (INIS)

2013-01-01

Seismic Safety Division of JNES is collecting and analyzing the findings of seismic safety research, and is developing a system to organize and disseminate the information internally and internationally. These tasks have been conducted in response to the lessons learned from Fukushima Daiichi NPP accident. The overview of the tasks is as follows; 1) Collection of the knowledge and findings from seismic safety research. JNES collects information on seismic safety researches including the 2011 off the Pacific coast of Tohoku Earthquake. The information is analyzed whether it is important for regulation to increase seismic safety of NPP. 2) Constructing database of seismic safety research. JNES collects information based on documents published by committee and constructs database of active faults around NPP sites in order to incorporate in the seismic safety review. 3) Dissemination of information related to seismic safety. JNES disseminates outcomes of own researches internally and internationally. (author)

Collection and accumulation of seismic safety research findings, and considerations for information dissemination

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Seismic Safety Division of JNES is collecting and analyzing the findings of seismic safety research, and is developing a system to organize and disseminate the information internally and internationally. These tasks have been conducted in response to the lessons learned from Fukushima Daiichi NPP accident. The overview of the tasks is as follows; 1) Collection of the knowledge and findings from seismic safety research. JNES collects information on seismic safety researches including the 2011 off the Pacific coast of Tohoku Earthquake. The information is analyzed whether it is important for regulation to increase seismic safety of NPP. 2) Constructing database of seismic safety research. JNES collects information based on documents published by committee and constructs database of active faults around NPP sites in order to incorporate in the seismic safety review. 3) Dissemination of information related to seismic safety. JNES disseminates outcomes of own researches internally and internationally. (author)

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

Nuclear emergency preparedness. Final report of the Nordic nuclear safety research project BOK-1

Energy Technology Data Exchange (ETDEWEB)

Lauritzen, Bent [Risoe National Lab., Roskilde (Denmark)

2002-02-01

Final report of the Nordic Nuclear Safety Research project BOK-1. The BOK-1 project, 'Nuclear Emergency Preparedness', was carried out in 1998-2001 with participants from the Nordic and Baltic Sea regions. The project consists of six sub-projects: Laboratory measurements and quality assurance (BOK-1.1); Mobile measurements and measurement strategies (BOK-1.2); Field measurement and data assimilation (BOK-1.3); Countermeasures in agriculture and forestry (BOK-1.4); Emergency monitoring in the Nordic and Baltic Sea countries (BOK-1.5); and Nuclear exercises (BOK-1.6). For each sub-project, the project outline, objectives and organization are described and main results presented. (au)

Nuclear emergency preparedness. Final report of the Nordic nuclear safety research project BOK-1

International Nuclear Information System (INIS)

Lauritzen, Bent

2002-02-01

Final report of the Nordic Nuclear Safety Research project BOK-1. The BOK-1 project, 'Nuclear Emergency Preparedness', was carried out in 1998-2001 with participants from the Nordic and Baltic Sea regions. The project consists of six sub-projects: Laboratory measurements and quality assurance (BOK-1.1); Mobile measurements and measurement strategies (BOK-1.2); Field measurement and data assimilation (BOK-1.3); Countermeasures in agriculture and forestry (BOK-1.4); Emergency monitoring in the Nordic and Baltic Sea countries (BOK-1.5); and Nuclear exercises (BOK-1.6). For each sub-project, the project outline, objectives and organization are described and main results presented. (au)

Development of safety-relevant components for the transport and handling of final storage casks for waste from decommissioning

International Nuclear Information System (INIS)

Bruening, D.; Geiser, H.; Kloeckner, F.; Rittscher, D.; Schlesinger, H.J.

1992-10-01

The aim of the study was the development, construction and testing of a transportation system that is able to transport cylindrical waste containers as well as containers from the deliverer to the 'KONRAD' final repository. A transport palette has been developed that can carry two cylindrical waste containers with type B requirement or classification II. An Open-All-Container for the transport of palettes and 'KONRAD' containers has been developed. A storage of cylindrical waste containers and containers in the final repository is possible with the newly developed transportation system. Safety specifications of the transportation system have been passed successfully. (orig.). 30 refs., 8 tabs., 74 figs [de

The safety leadership.

Science.gov (United States)

Lees, Howard; Faulkner, Bruce

2010-01-01

In this paper the authors present a careful consideration about the role of leadership, the fundamental element for the success of Behavior-Based Safety (B-BS) programs within companies. Lees and Faulkner have been training, coaching and writing about Behavior-Based Safety for the last ten years. Considerable data has been gathered during this process and the paramount factor in its success is leadership. An effective leader can create many spectacular successes. The success stories are all predicated on good leadership, without that a good product, great processes and quality people are all wasted and often find themselves on the rocks of frustration.

Site safety progress review of spent fuel central interim storage facility. Final report

International Nuclear Information System (INIS)

Gurpinar, A.; Serva, L.; Giuliani

1995-01-01

Following the request of the Czech Power Board (CEZ) and within the scope of the Technical Cooperation Project CZR/9/003, a progress review of the site safety of the Spent Fuel Central Interim Storage Facility (SFCISF) was performed. The review involved the first two stages of the works comprising the regional survey and identification of candidate sites for the underground and surface storage options. Five sites have been identified as a result of the previous works. The following two stages will involved the identification of the preferred candidate sites for the two options and the final site qualification. The present review had the purpose of assessing the work already performed and making recommendations for the next two stages of works

Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

Energy Technology Data Exchange (ETDEWEB)

Bollingerfehr, W.; Buhmann, D.; Doerr, S.; and others

2017-03-15

safety demonstration are the integrity proofs for the geological and geotechnical barriers and analysis of backfill compaction. In addition, any possible radionuclide release from the repository to the environment has also to be assessed. The safety and demonstration concept developed in the course of the ISIBEL project was further evolved and applied in the course of the R and D project ''Vorlaeufige Sicherheitsanalyse Gorleben - VSG'' (preliminary safety analysis Gorleben) as an example for an HLW repository in a domal salt structure. The repository concepts also consider the requirement for retrievability of stored waste during the operational phase of the repository. The results of the R and D project VSG provide evidence that a safe HLW repository within a salt dome of a suitable geologic structure is feasible. The long-term safety can be ensured using state-of-the-art science and technology. In 2010, the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) issued new safety requirements for the disposal of heat-generating radioactive waste. These requirements have been included in the analysis. After completion of the VSG project in 2013 complementary work has been performed within the framework of the ISIBEL programme. In this context e.g. potential contributions of natural and antropogenic analogs to confidence building were addressed as well as the feasibility and limits of deriving a repository conc ept strictly from requirements. The report in hands provides a comprehensive summary of the results of R and D work regarding HLW disposal in domal salt formations that has been performed after launching the ISIBEL programme in 2005. This study shows the depth of the geological and technical knowledge on final disposal of HLW in a salt dome with a suitable geologic structure that had been gained up to now and demonstrates that the tools required for safety evaluations are available and allow reliable safety

Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

International Nuclear Information System (INIS)

Bollingerfehr, W.; Buhmann, D.; Doerr, S.

2017-03-01

safety demonstration are the integrity proofs for the geological and geotechnical barriers and analysis of backfill compaction. In addition, any possible radionuclide release from the repository to the environment has also to be assessed. The safety and demonstration concept developed in the course of the ISIBEL project was further evolved and applied in the course of the R and D project ''Vorlaeufige Sicherheitsanalyse Gorleben - VSG'' (preliminary safety analysis Gorleben) as an example for an HLW repository in a domal salt structure. The repository concepts also consider the requirement for retrievability of stored waste during the operational phase of the repository. The results of the R and D project VSG provide evidence that a safe HLW repository within a salt dome of a suitable geologic structure is feasible. The long-term safety can be ensured using state-of-the-art science and technology. In 2010, the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) issued new safety requirements for the disposal of heat-generating radioactive waste. These requirements have been included in the analysis. After completion of the VSG project in 2013 complementary work has been performed within the framework of the ISIBEL programme. In this context e.g. potential contributions of natural and antropogenic analogs to confidence building were addressed as well as the feasibility and limits of deriving a repository conc ept strictly from requirements. The report in hands provides a comprehensive summary of the results of R and D work regarding HLW disposal in domal salt formations that has been performed after launching the ISIBEL programme in 2005. This study shows the depth of the geological and technical knowledge on final disposal of HLW in a salt dome with a suitable geologic structure that had been gained up to now and demonstrates that the tools required for safety evaluations are available and allow reliable safety assessments of HLW

Discussion on building safety culture inside a nuclear safety regulatory body

International Nuclear Information System (INIS)

Fan Yumao

2013-01-01

A strong internal safety culture plays a key role in improving the performance of a nuclear regulatory body. This paper discusses the definition of internal safety culture of nuclear regulatory bodies, and explains the functions that the safety culture to facilitate the nuclear safety regulation and finally puts forward some thoughts about building internal safety culture inside regulatory bodies. (author)

Safety margins in deterministic safety analysis

International Nuclear Information System (INIS)

Viktorov, A.

2011-01-01

The concept of safety margins has acquired certain prominence in the attempts to demonstrate quantitatively the level of the nuclear power plant safety by means of deterministic analysis, especially when considering impacts from plant ageing and discovery issues. A number of international or industry publications exist that discuss various applications and interpretations of safety margins. The objective of this presentation is to bring together and examine in some detail, from the regulatory point of view, the safety margins that relate to deterministic safety analysis. In this paper, definitions of various safety margins are presented and discussed along with the regulatory expectations for them. Interrelationships of analysis input and output parameters with corresponding limits are explored. It is shown that the overall safety margin is composed of several components each having different origins and potential uses; in particular, margins associated with analysis output parameters are contrasted with margins linked to the analysis input. While these are separate, it is possible to influence output margins through the analysis input, and analysis method. Preserving safety margins is tantamount to maintaining safety. At the same time, efficiency of operation requires optimization of safety margins taking into account various technical and regulatory considerations. For this, basic definitions and rules for safety margins must be first established. (author)

Comparative assessment of nanomaterial definitions and safety evaluation considerations.

Science.gov (United States)

Boverhof, Darrell R; Bramante, Christina M; Butala, John H; Clancy, Shaun F; Lafranconi, Mark; West, Jay; Gordon, Steve C

2015-10-01

Nanomaterials continue to bring promising advances to science and technology. In concert have come calls for increased regulatory oversight to ensure their appropriate identification and evaluation, which has led to extensive discussions about nanomaterial definitions. Numerous nanomaterial definitions have been proposed by government, industry, and standards organizations. We conducted a comprehensive comparative assessment of existing nanomaterial definitions put forward by governments to highlight their similarities and differences. We found that the size limits used in different definitions were inconsistent, as were considerations of other elements, including agglomerates and aggregates, distributional thresholds, novel properties, and solubility. Other important differences included consideration of number size distributions versus weight distributions and natural versus intentionally-manufactured materials. Overall, the definitions we compared were not in alignment, which may lead to inconsistent identification and evaluation of nanomaterials and could have adverse impacts on commerce and public perceptions of nanotechnology. We recommend a set of considerations that future discussions of nanomaterial definitions should consider for describing materials and assessing their potential for health and environmental impacts using risk-based approaches within existing assessment frameworks. Our intent is to initiate a dialogue aimed at achieving greater clarity in identifying those nanomaterials that may require additional evaluation, not to propose a formal definition. Copyright © 2015 The Authors. Published by Elsevier Inc. All rights reserved.

Defence-in-depth strategy of fire protection and its relevance after final shutdown (by the example of Germany)

Energy Technology Data Exchange (ETDEWEB)

Beesen, Michael; Ernst, Benjamin; Fischer, Guenter [TUeV SUeD Industrie Service GmbH (Germany)

2015-12-15

Nuclear power plants (NPP) are protected against internal and external fires by a fire protection defence-in-depth concept including the following precautionary measures: operational, structural and equipment related fire protection measures as well as manual fire fighting. The fire protection measures are designed in consideration of fires to be expected (from fire loads permanently and temporarily present together with potential ignition sources) in order to prevent a violation of both the protection goals of public law and the nuclear protection goals / radiological safety objectives in case of internal and external fires. The aspect ''What is the future significance of the fire protection defence-in-depth concept?'' needs to be considered with regard to the situation following the final shutdown. From our point of view as a TSO (technical safety organization) both the non-nuclear protection goals (e.g. prevent occurrence of a fire; ensure escape and rescue of humans) as well as the nuclear ones have to be ensured after final shutdown of a nuclear plant. The protection goals of public law will almost completely remain after the plant has stopped commercial operation while the nuclear safety objectives will be stepwise reduced in consideration of the decommissioning status until the end of the nuclear supervision. Nevertheless, the fire protection concept must clearly specify those fire protection measures that are necessary to ensure the plants' safety. The situation on site regularly needs to be under examination to check if the fire protection concept covers all conditions to be considered and if the existing fire protection measures are sufficient or if an adaption is necessary.

Safety and protection aspects of the management of high-level radioactive wastes

International Nuclear Information System (INIS)

Candes, P.; Pradel, J.

1977-01-01

Appropriate consideration is given in France to safety and protection problems to be solved from production up to the final disposal of high-level radioactive wastes. The first stage of this work consisted in emphasizing the various technical options. Different strategies appear to be possible, taking into account the technical, political and psychological difficulties. This results in evaluating the safety problems to be solved in the framework of those strategies. In this field the main safety and protection principles do not differ from those applying to other nuclear facilities. Nevertheless, the factor of time is in most cases quite different (thousands or millions of years). The question is then raised of evaluating the importance to be given to very remote consequences, both at philosophical and at scientific levels. As a first result of these considerations, the application of the barrier concept is recommended. This concept is familiar to safety specialists. Different barriers, for which particular problems are listed and evaluated, are defined. Another result with regard to radiation protection principles is to consider that if safety provisions should lead to as efficient a containment of radioactive products as possible, it would not be realistic to consider such a containment as absolute, in particular for disposal lasting anything up to thousands of years. It is therefore assumed that a limited radioactivity transfer should be taken into account, and its consequences for the environment and man calculated. This is especially true in the study of an appropriate site for final storage, and the study should necessarily include a detailed investigation of the retention characteristics of soil layers, and the implementation of appropriate models giving a sufficiently accurate evaluation of the consequences of transfers, including those related to the effect of various elements after their arrival into the biosphere. The authors review the different

Safety assessment input for site selection - the Swedish example - 59031

International Nuclear Information System (INIS)

Andersson, Johan

2012-01-01

the calculated risk for Forsmark will be considerably lower than that for Laxemar.. What decided the choice is that Forsmark is thus mainly that it was judged to offer better prospects for achieving long-term safety in the final repository. Other factors considered included implications regarding repository construction and operation, difference in the footprint of the repository, comparisons regarding environment and health as well as social resources and local support. (authors)

Predisposal Management of Low and Intermediate Level Radioactive Waste. Safety Guide

International Nuclear Information System (INIS)

2009-01-01

The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established for the predisposal management of low and intermediate level waste. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. General safety considerations; 5. Safety features for the predisposal management of LILW; 6. Record keeping and reporting; 7. Safety assessment; 8. Quality assurance; Annex I: Nature and sources of LILW from nuclear facilities; Annex II: Development of specifications for waste packages; Annex III: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Postulated initiating events for consideration in a safety assessment (internal phenomena).

77 FR 24560 - National Highway Traffic Safety Administration Electric Vehicle Safety Technical Symposium

Science.gov (United States)

2012-04-24

... DEPARTMENT OF TRANSPORTATION National Highway Traffic Safety Administration Electric Vehicle... discuss safety considerations for electric vehicles powered by lithium-ion (Li-ion) batteries. The... vehicles. Electric vehicles show great promise as an innovative and fuel- efficient option for American...

Non-clinical models: validation, study design and statistical consideration in safety pharmacology.

Science.gov (United States)

Pugsley, M K; Towart, R; Authier, S; Gallacher, D J; Curtis, M J

2010-01-01

The current issue of the Journal of Pharmacological and Toxicological Methods (JPTM) focuses exclusively on safety pharmacology methods. This is the 7th year the Journal has published on this topic. Methods and models that specifically relate to methods relating to the assessment of the safety profile of a new chemical entity (NCE) prior to first in human (FIH) studies are described. Since the Journal started publishing on this topic there has been a major effort by safety pharmacologists, toxicologists and regulatory scientists within Industry (both large and small Pharma as well as Biotechnology companies) and also from Contract Research Organizations (CRO) to publish the surgical details of the non-clinical methods utilized but also provide important details related to standard and non-standard (or integrated) study models and designs. These details from core battery and secondary (or ancillary) drug safety assessment methods used in drug development programs have been the focus of these special issues and have been an attempt to provide validation of methods. Similarly, the safety pharmacology issues of the Journal provide the most relevant forum for scientists to present novel and modified methods with direct applicability to determination of drug safety-directly to the safety pharmacology scientific community. The content of the manuscripts in this issue includes the introduction of additional important surgical methods, novel data capture and data analysis methods, improved study design and effects of positive control compounds with known activity in the model. Copyright 2010 Elsevier Inc. All rights reserved.

Lead plant application of leak-before-break to high energy piping. Final report, January 1989

International Nuclear Information System (INIS)

1989-01-01

This report presents the experience gained during a successful application of a leak-before-break program by Duquesne Light Company. This program was directed at the high energy nuclear piping at Beaver Valley Power Station - Unit 2. This experience can be applied to other nuclear plant leak-before-break efforts in order to minimize the number of pipe whip restraints, jet impingement shields, snubbers, and to discount the consideration of remaining pipe rupture dynamic effects. The chronology of events leading to Nuclear Regulatory Commission approval of the Beaver Valley Power Station - Unit 2 lead plant effort is described. The final report and pertinent sections of the final Safety Evaluation Report are also included. (author)

DOE Defense Program (DP) safety programs. Final report, Task 003

International Nuclear Information System (INIS)

1998-01-01

The overall objective of the work on Task 003 of Subcontract 9-X52-W7423-1 was to provide LANL with support to the DOE Defense Program (DP) Safety Programs. The effort included the identification of appropriate safety requirements, the refinement of a DP-specific Safety Analysis Report (SAR) Format and Content Guide (FCG) and Comprehensive Review Plan (CRP), incorporation of graded approach instructions into the guidance, and the development of a safety analysis methodologies document. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective is provided here

EU DEMO blanket concepts safety assessment. Final report of Working Group 6a of the Blanket Concept Selection Exercise

International Nuclear Information System (INIS)

Kleefeldt, K.; Porfiri, T.

1996-06-01

The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four blanket concepts under development. Two of them use lithium ceramics, the other two concepts employ an eutectic lead-lithium alloy (Pb-17Li) as breeder material. The two most promising concepts were to select in 1995 for further development. In order to prepare the selection, a Blanket Concept Selection Exercise (BCSE) has been inititated by the participating associations under the auspices of the European Commission. This BCSE has been performed in 14 working groups which, in a comparative evaluation of the four blanket concepts, addressed specific fields. The working group safety addressed the safety implications. This report describes the methodology adopted, the safety issues identified, their comparative evaluation for the four concepts, and the results and conclusions of the working group to be entered into the overall evaluation. There, the results from all 14 working groups have been combined to yield a final ranking as a basis for the selection. In summary, the safety assessment showed that the four European blanket concepts can be considered as equivalent in terms of the safety rating adopted, each concept, however, rendering safety concerns of different quality in different areas which are substantiated in this report. (orig.) [de

Information need about the safety of the final disposal of nuclear waste. Information receiver`s views in Eurajoki, Kuhmo and Aeaenekoski municipalities; Tiedontarve ydinjaetteen loppusijoituksen turvallisuudesta. Vastaanottajan naekoekulmia Eurajoella, Kuhmossa ja Aeaenekoskella

Energy Technology Data Exchange (ETDEWEB)

Hautakangas, H

1997-03-01

The study analyses the public`s information need about the safety issues related to the final disposal of spent nuclear fuel generated by the Finnish nuclear power stations. Locals in three municipalities that are studied as possible sites for final disposal were interviewed for the study. Earlier studies made in Finland had indicated that the public`s knowledge about safety issues related to the final disposal was almost opposite to the findings of the natural sciences. Also, the public had expressed a wish to receive more information from the safety authority, the Finnish Centre for Radiation and Nuclear Safety (STUK). This study therefore had two basic objectives: To find out what kind of safety information the locals need and what the safety authority`s role could be in providing information. The main results show interest and need especially for information concerning the disposal phases taking place on the ground level, such as nuclear waste transportation and encapsulation. Also, the interviews show a clear need and desire for an impartial actor such as STUK in the information and communication process. (author) (107 refs.).

Criticality safety considerations. Integral Monitored Retrievable Storage (MRS) Facility

International Nuclear Information System (INIS)

1986-09-01

This report summarizes the criticality analysis performed to address criticality safety concerns and to support facility design during the conceptual design phase of the Monitored Retrievable Storage (MRS) Facility. The report addresses the criticality safety concerns, the design features of the facility relative to criticality, and the results of the analysis of both normal operating and hypothetical off-normal conditions. Key references are provided (Appendix C) if additional information is desired by the reader. The MRS Facility design was developed and the related analysis was performed in accordance with the MRS Facility Functional Design Criteria and the Basis for Design. The detailed description and calculations are documented in the Integral MRS Facility Conceptual Design Report. In addition to the summary portion of this report, explanatary notes for various terms, calculation methodology, and design parameters are presented in Appendix A. Appendix B provides a brief glossary of technical terms

Appendix C: safety design rationale

International Nuclear Information System (INIS)

Ghose, S.

1985-01-01

A brief discussion of the rationale for safety design of fusion plants is presented in the main text. Further detail safety considerations are presented in this appendix in the form of charts and tables. The author present some of the major safety criteria and other criteria used in blanket selection here

The nuclear fuel cycle including essential aspects of safety

International Nuclear Information System (INIS)

Warnemuende, R.; May, H.

1978-11-01

When judging nuclear energy not only the reactor but also the whole fuel cycle is of importance. The fuel cycle consists of the supply, i.e. the process from uranium ore to the insertion of fuel elements into the reactor and the waste management, the removal of fuel elements from the reactor and the final storage of radioactive waste. The different stages of the nuclear fuel cycle are well known with regard to their technical difficulties, their problems of industrial safety and pollution. Although it is possible to compare them qualitatively, they still differ partly to a considerable extent, from a quantitative point of view. However, the fact that technical solutions are available for all kinds of tasks can be stated. It is significant for the Federal Republic of Germany that all essential preparatory work for closing the nuclear fuel cycle has been carried out and that safety problems will no longer be in the way of the large-scale realization of uranium enrichment, reprocessing of nuclear fuels and final storage of radioactive waste. Further research and development activities will serve its technical and economic optimization. (orig.) [de

Sixth national report of Brazil for the nuclear safety convention

International Nuclear Information System (INIS)

2013-01-01

Brazil has presented periodically its National Report prepared by a group composed of representatives of the various Brazilian organizations with responsibilities related to nuclear safety. Due to the implications of the Fukushima nuclear accident in 2011, an Extraordinary National Report was presented in 2012. This Sixth National Report is an update of the Fifth National Report in relation to the Convention on Nuclear Safety articles and also an update of the Extraordinary Report with respect to the action taken related to lesson learned from the Fukushima accident. It includes relevant information for the period of 2010/2012. This document represents the national report prepared as a fulfillment of the brazilian obligations related to the Convention on Nuclear Safety. In chapter 2 some details are given about the existing nuclear installations. Chapter 3 provides details about the legislation and regulations, including the regulatory framework and the regulatory body. Chapter 4 covers general safety considerations as described in articles 10 to 16 of the Convention. Chapter 5 addresses to the safety of the installations during siting, design, construction and operation. Chapter 6 describes planned activities to further enhance nuclear safety. Chapter 7 presents the final remarks related to the degree of compliance with the Convention obligations

Sixth national report of Brazil for the nuclear safety convention

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-07-01

Brazil has presented periodically its National Report prepared by a group composed of representatives of the various Brazilian organizations with responsibilities related to nuclear safety. Due to the implications of the Fukushima nuclear accident in 2011, an Extraordinary National Report was presented in 2012. This Sixth National Report is an update of the Fifth National Report in relation to the Convention on Nuclear Safety articles and also an update of the Extraordinary Report with respect to the action taken related to lesson learned from the Fukushima accident. It includes relevant information for the period of 2010/2012. This document represents the national report prepared as a fulfillment of the brazilian obligations related to the Convention on Nuclear Safety. In chapter 2 some details are given about the existing nuclear installations. Chapter 3 provides details about the legislation and regulations, including the regulatory framework and the regulatory body. Chapter 4 covers general safety considerations as described in articles 10 to 16 of the Convention. Chapter 5 addresses to the safety of the installations during siting, design, construction and operation. Chapter 6 describes planned activities to further enhance nuclear safety. Chapter 7 presents the final remarks related to the degree of compliance with the Convention obligations.

Design safety improvements of Kozloduy NPP to meet the modern safety requirements towards the old generation PWR

International Nuclear Information System (INIS)

Hinovski, M.P.; Sabinov, S.

2001-01-01

Activities related to safety improvement of Kozloduy NPP units, started at the end of 1970s included seismic resistance upgrading, fire safety improvement, reliable heat final absorber etc. During the last 10 years the approach was systematized and improved. Units 1 to 4 are of great interest; therefore here we will discuss these units only. As a result of studies and analyses performed at the end of the 1980s and the beginning of the 1990s, problems related to the safety were identified and complex of technical measures was developed and planned. A considerable part of these measures has already been implemented, and the rest will be performed during the next years. Activities were performed by stages, and at the moment the last stage is under way. It shall be finished by the year 2003. The number of the measures is quite large to describe them here in full scope -- during the first stage of the safety program (1991-1993) were developed and analyzed more than 4200 documents and more than 160 measures were executed. During the second and third stages more than 300 important improvements were realized. In the frame of the program, financed by EBRD, 10 new systems with great importance were implemented and 8 systems were significantly modified. The main measures are described below. (author)

Long Duration Head-Down Tilt Bed Rest Studies: Safety Considerations Regarding Vision Health

Science.gov (United States)

Cromwell, Ronita L.; Zanello, S. B.; Yarbough, P. O.; Ploutz-Snyder, Robert; Taibbi, G.; Vizzeri, G.

2012-01-01

Visual symptoms reported in astronauts returning from long duration missions in low Earth orbit, including hyperopic shift, choroidal folds, globe flattening and papilledema, are thought to be related to fluid shifts within the body due to microgravity exposure. Because of this possible relation to fluid shifts, safety considerations have been raised regarding the ocular health of head-down tilt (HDT) bed rest subjects. HDT is a widely used ground ]based analog that simulates physiological changes of spaceflight, including fluid shifts. Thus, vision monitoring has been performed in bed rest subjects in order to evaluate the safety of HDT with respect to vision health. Here we report ocular outcomes in 9 healthy subjects (age range: 27-48 years; Male/Female ratio: 8/1) completing bed rest Campaign 11, an integrated, multidisciplinary 70-day 6 degrees HDT bed rest study. Vision examinations were performed on a weekly basis, and consisted of office-based (2 pre- and 2 post-bed rest) and in-bed testing. The experimental design was a repeated measures design, with measurements for both eyes taken for each subject at each planned time point. Findings for the following tests were all reported as normal in each testing session for every subject: modified Amsler grid, red dot test, confrontational visual fields, color vision and fundus photography. Overall, no statistically significant differences were observed for any of the measures, except for both near and far visual acuity, which increased during the course of the study. This difference is not considered clinically relevant as may result from the effect of learning. Intraocular pressure results suggest a small increase at the beginning of the bed rest phase (p=0.059) and lesser increase at post-bed rest with respect to baseline (p=0.046). These preliminary results provide the basis for further analyses that will include correlations between intraocular pressure change pre- and post-bed rest, and optical coherence

The role of staff training in the safety of nuclear facilities

International Nuclear Information System (INIS)

Koechlin, J.C.; Tanguy, P.

1980-01-01

Although nuclear energy largely involves automatic protection systems enabling the effects of human error to be mitigated, the human factor still remains of extreme importance in nuclear risk analysis. Hence, the attainment of the high safety standards sought after for nuclear energy must of necessity entail staff training programmes which take into account the concern for nuclear safety. It is incumbent upon constructors and operators to evolve a training programme suited to each job, and the safety authorities are responsible for assessing whether the programme is satisfactory from the standpoint of safety and, where necessary, for issuing the relevant certificates or permits. The paper makes some comments on the cost of human error and the profitability of investment in training, on the importance of practical training and of the role of simulators, and on the need for operators to note and analyse all operational abnormalities, which are so often an advance warning of accidents. The training of special safety teams is examined, with consideration of three aspects: safety assessment, inspection, and action to be taken in the event of accident. Finally, some information is given on the human reliability studies under way and their implications for nuclear safety and training, with emphasis on the valuable assistance rendered in this matter by international organizations. (author)

Basic Program Elements for Federal employee Occupational Safety and Health Programs and related matters; Subpart I for Recordkeeping and Reporting Requirements. Final rule.

Science.gov (United States)

2013-08-05

OSHA is issuing a final rule amending the Basic Program Elements to require Federal agencies to submit their occupational injury and illness recordkeeping information to the Bureau of Labor Statistics (BLS) and OSHA on an annual basis. The information, which is already required to be created and maintained by Federal agencies, will be used by BLS to aggregate injury and illness information throughout the Federal government. OSHA will use the information to identify Federal establishments with high incidence rates for targeted inspection, and assist in determining the most effective safety and health training for Federal employees. The final rule also interprets several existing basic program elements in our regulations to clarify requirements applicable to Federal agencies, amends the date when Federal agencies must submit to the Secretary of Labor their annual report on occupational safety and health programs, amends the date when the Secretary of Labor must submit to the President the annual report on Federal agency safety and health, and clarifies that Federal agencies must include uncompensated volunteers when reporting and recording occupational injuries and illnesses.

Adoptable Interventions, Human Health, and Food Safety Considerations for Reducing Sodium Content of Processed Food Products

Science.gov (United States)

Allison, Abimbola; Fouladkhah, Aliyar

2018-01-01

Although vital for maintaining health when consumed in moderation, various epidemiological studies in recent years have shown a strong association between excess dietary sodium with an array of health complications. These associations are robust and clinically significant for development of hypertension and prehypertension, two of the leading causes of preventable mortality worldwide, in adults with a high-sodium diet. Data from developed nations and transition economies show worldwide sodium intake of higher than recommended amounts in various nations. While natural foods typically contain a moderate amount of sodium, manufactured food products are the main contributor to dietary sodium intake, up to 75% of sodium in diet of American adults, as an example. Lower cost in formulation, positive effects on organoleptic properties of food products, effects on food quality during shelf-life, and microbiological food safety, make sodium chloride a notable candidate and an indispensable part of formulation of various products. Although low-sodium formulation of each product possesses a unique set of challenges, review of literature shows an abundance of successful experiences for products of many categories. The current study discusses adoptable interventions for product development and reformulation of products to achieve a modest amount of final sodium content while maintaining taste, quality, shelf-stability, and microbiological food safety. PMID:29389843

Adoptable Interventions, Human Health, and Food Safety Considerations for Reducing Sodium Content of Processed Food Products.

Science.gov (United States)

Allison, Abimbola; Fouladkhah, Aliyar

2018-02-01

Although vital for maintaining health when consumed in moderation, various epidemiological studies in recent years have shown a strong association between excess dietary sodium with an array of health complications. These associations are robust and clinically significant for development of hypertension and prehypertension, two of the leading causes of preventable mortality worldwide, in adults with a high-sodium diet. Data from developed nations and transition economies show worldwide sodium intake of higher than recommended amounts in various nations. While natural foods typically contain a moderate amount of sodium, manufactured food products are the main contributor to dietary sodium intake, up to 75% of sodium in diet of American adults, as an example. Lower cost in formulation, positive effects on organoleptic properties of food products, effects on food quality during shelf-life, and microbiological food safety, make sodium chloride a notable candidate and an indispensable part of formulation of various products. Although low-sodium formulation of each product possesses a unique set of challenges, review of literature shows an abundance of successful experiences for products of many categories. The current study discusses adoptable interventions for product development and reformulation of products to achieve a modest amount of final sodium content while maintaining taste, quality, shelf-stability, and microbiological food safety.

Safety culture: modern slogan or effective contribution to safety?

International Nuclear Information System (INIS)

Salm, M.

1994-01-01

Safety culture is defined and its impact on nuclear power plants is documented using the words of the INSAG of IAEA. Two examples from the field of aviation and space flight testify, that the upper management, by its sheer image, may considerably influence actions of the lower levels of the hierarchy. Management therefore can do a lot more for safety than is commonly assumed. Two examples, although separated by 57 years, show that the mentioned influence remains unchanged inspire of progress in management- and organisation-methods as well as in safety-engineering. Safety culture is an overriding element of safety, acting at all levels of a hierarchy. Its action is most important on those levels, for which precise reglementation is hardly possible. The chain of technical and organisational measures guarantees safety only under the condition, that it is embedded in 'safety culture'. Safety culture therefore merits our full attention. (author) 1 fig

AEC sets five year nuclear safety research program

International Nuclear Information System (INIS)

Anon.

1976-01-01

The research by the government for the establishment of means of judging the adequacy of safety measures incorporated in nuclear facilities, including setting safety standards and collecting documents of general criteria, and the research by the industry on safety measures and the promotion of safety-related technique are stated in the five year program for 1976-80 reported by subcommittees, Atomic Energy Commission (AEC). Four considerations on the research items incorporated in the program are 1) technical programs relating to the safety of nuclear facilities and the necessary criteria, 2) priority of the relevant items decided according to their impact on circumstances, urgency, the defence-indepth concept and so on, 3) consideration of all relevant data and documents collected, and research subjects necessary to quantify safety measurement, and 4) consideration of technological actualization, the capability of each research body, the budget and the time schedule. In addition, seven major themes decided on the basis of these points are 1) reactivity-initiated accident, 2) LOCA, 3) fuel behavior, 4) structural safety, 5) radioactive release, 6) statistical method of safety evaluation, and 7) seismic characteristics. The committee has deliberated the appropriate division of researches between the government and the industry. A set of tables showing the nuclear safety research plan for 1976-80 are attached. (Iwakiri, K.)

Exploration and safety evaluations of salt formations and site selection procedures; Erkundung und Sicherheitsbewertung von Salzformationen und Standortauswahlverfahren

Energy Technology Data Exchange (ETDEWEB)

Krapf, Eva Barbara

2016-12-12

In 2011 the final decision for the withdrawal from the nuclear energy program was decided in the Federal Republic of Germany. The majority of the produced radioactive waste originate in the operation as well as in the decommissioning and dismantling of nuclear facilities. The long-term containment of especially heat-developing and high-level waste in an underground disposal facility is pursued. The Site Selection Act (StandAG), passed in 2013, defined further procedural steps as well as responsibilities and the way of public participation during the site selection. In this context the newly founded Commission Storage of Highly Radioactive Waste was assigned with the task of giving relevant recommendations based on their investigation of specific aspects and fundamental questions. The objective of this procedure is the selection of the site that can provide the best possible safety for humans and the environment during the defined period of one million years. The Commissions' final report was published in July 2016. In this thesis a possible approach for exploring sites in connection with safety investigations is recommended. The site selection procedure described in the StandAG represents the basis for the considerations. Geoscientific exclusion criteria, minimum requirements as well as weighing criteria can be developed regarding the relevant geoscientific and climatic changes during the defined period of one million years. In contrast to the recommendations made by the Commission Storage of Highly Radioactive Waste no previously existing report has been revised and adapted. Rather, all issues relevant for the long-term containment of radioactive waste in a disposal facility had been newly developed. The considerations are related to salt domes as host rock. Furthermore, according to the StandAG preliminary safety investigations are required in every step of the site selection. The recommendations made in this thesis concerning content and feasibility of

The safety of nuclear power: Strategy for the future

International Nuclear Information System (INIS)

1992-01-01

The conference took place in Vienna from 2 to 6 September 1991. It was attended by approximately 350 participants from about fifty countries and 12 international organizations. The conference was directed to decision makers on nuclear safety and energy policy at the technical policy level. Its objective was to review the nuclear power safety issues on which international consensus would be desirable, to address the concerns on nuclear safety expressed by the WCED, and to formulate recommendations for future actions by national and international authorities to advance nuclear safety to the highest level, including proposals for the IAEA's future activities for consideration by its governing bodies. Background Papers were prepared in advance of the conference by Expert Groups on the following five issues: Fundamental principles for the safe use of nuclear power; Ensuring and enhancing safety of operating plants; Treatment of nuclear power plants built to earlier safety standards; The next generation of nuclear power plants; Final disposal of radioactive waste. On the basis of comments received on these papers from IAEA Member States, significant topics for discussion were identified. These topics and the papers formed the basis of the discussions from which the conference arrived at recommendations for future action by national and international authorities. A separate abstract was prepared for the opening speeches, background papers, major findings of the conference and the President's closing statement. 2 figs, 1 tab

Preliminary considerations on safety of computerized control rooms

International Nuclear Information System (INIS)

Vittet, J.

1983-02-01

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

Safety and effectiveness considerations for clinical studies of visual prosthetic devices

Science.gov (United States)

Cohen, Ethan D.

2007-03-01

With the advent of new designs of visual prostheses for the blind, FDA is faced with developing guidance for evaluating their engineering, safety and patient performance. Visual prostheses are considered significant risk medical devices, and their use in human clinical trials must be approved by FDA under an investigation device exemption (IDE). This paper contains a series of test topics and design issues that sponsors should consider in order to assess the safety and efficacy of their device. The IDE application includes a series of pre-clinical and clinical data sections. The pre-clinical section documents laboratory, animal and bench top performance tests of visual prostheses safety and reliability to support a human clinical trial. The materials used in constructing the implant should be biocompatible, sterile, corrosion resistant, and able to withstand any forces exerted on it during normal patient use. The clinical data section is composed of items related to patient-related evaluation of device performance. This section documents the implantation procedure, trial design, statistical analysis and how visual performance is assessed. Similar to cochlear implants, a visual prosthesis is expected to last in the body for many years, and good pre-clinical and clinical testing will help ensure its safety, durability and effectiveness.

Financial Considerations for Health and Safety in the Australian Dairy Industry.

Science.gov (United States)

Lower, Tony; Pollock, Kirrily

2017-01-01

This study aimed to quantify the burden of fatal and workers' compensation-related injuries on the dairy sector in Australia and to assess current financial inputs for work health and safety actions. The first and second phases entailed a review of the financial costs associated with fatal (2001-2013) and workers' compensation cases (2008-2009 to 2012-2013). The third element was a case-study approach focusing on actual production and cost data from a small sample of dairy farms (n = 6) to ascertain an estimate for current financial inputs to work health and safety. The estimated financial impact was approximately AUD$12.4 million each year, comprising AUD$6.8 million for fatalities and AUD$5.6 million for compensable injuries. All sample businesses were investing in health and safety initiatives; however, this was modest and on average represented 0.4% of total expenditure (AUD$4,472) or financial case to increase strategic investments into work health and safety by dairy farmers in Australia. Increased investment has the potential to not only minimize the burden of injury and costs to dairy farmers, but will also significantly reduce the devastating impacts that these incidents have on individuals, their families and communities.

Developing international safety standards for the geological disposal of radioactive waste

International Nuclear Information System (INIS)

Metcalf, P.

2001-01-01

In the context of the International Atomic Energy Agency's (IAEA) programme to create a corpus of internationally accepted Radioactive Waste Safety Standards (RADWASS), focus is currently being placed on establishing standards for the 'geological disposal of radioactive waste'. This is a challenging task and to help the standards development process there is a need to stimulate discussion of some of the associated scientific and technical issues. A number of position papers developed in recent years by a subgroup of the Waste Safety Standards Committee (WASSC), the subgroup on Principles and Criteria for Radioactive Waste Disposal, address many of the relevant issues. These include a common safety based framework for radioactive waste disposal, appropriate time frames for safety assessment, different possible indicators of long-term safety, the safety implications of reversibility and retrievability, the assessment of possible human intrusion into the repository, the role and limitations of institutional control, establishing reference critical groups and biospheres for long-term assessment, and what is meant by 'compliance' with the standards. These papers will be discussed at a Specialists Meeting to be held at the IAEA in June 2001 as a means of establishing the extent to which they enjoy the general support of experts. In order to broaden that consensus, the conclusions reached at the Specialists Meeting on the issues listed above will be presented and discussed with participants at a number of international meetings. Later this year, a draft safety standard on the geological disposal of radioactive waste which takes account of the consensus positions reached through the various consultations will be submitted for the consideration of Waste Safety Standards Committee (WASSC), the officially approved body within the IAEA for the review and approval of waste safety standards. The Committee is made up of government appointed radioactive waste regulators

Some considerations on the Tlatelolco Treaty

International Nuclear Information System (INIS)

Faria, N.M. de; Goes Fischer, M.D. de

1981-01-01

Some considerations related to the Tlatelolco Treaty are focused and so are the role and the position of Brazil in view of the Treaty. Short historical remarks are presented in order to show the commitments of Brazil with the Treaty. Finally, considerations concerned with the validity of the Treaty as the legal instrument to contribute to the security and peace in the World. (Author) [pt

20 CFR 405.373 - Requesting consideration of new evidence.

Science.gov (United States)

2010-04-01

... consideration of new evidence. (a) If the administrative law judge's decision is our final decision, the... consideration by the administrative law judge at the earliest possible opportunity, but no later than 30 days... 20 Employees' Benefits 2 2010-04-01 2010-04-01 false Requesting consideration of new evidence. 405...

Safety performance evaluation of converging chevron pavement markings : final report.

Science.gov (United States)

2014-12-01

The objectives of this study were (1) to perform a detailed safety analysis of converging chevron : pavement markings, quantifying the potential safety benefits and developing an understanding of the : incident types addressed by the treatment, and (...

Criticality safety in high explosives dissolution

International Nuclear Information System (INIS)

Troyer, S.D.

1997-01-01

In 1992, an incident occurred at the Pantex Plant in which the cladding around a fissile material component (pit) cracked during dismantlement of the high explosives portion of a nuclear weapon. Although the event did not result in any significant contamination or personnel exposures, concerns about the incident led to the conclusion that the current dismantlement process was unacceptable. Options considered for redesign, dissolution tooling design considerations, dissolution tooling design features, and the analysis of the new dissolution tooling are summarized. The final tooling design developed incorporated a number of safety features and provides a simple, self-contained, low-maintenance method of high explosives removal for nuclear explosive dismantlement. Analyses demonstrate that the tooling design will remain subcritical under normal, abnormal, and credible accident scenarios. 1 fig

Ferrocyanide tank safety program: Cesium uptake capacity of simulated ferrocyanide tank waste. Final report

International Nuclear Information System (INIS)

Burgeson, I.E.; Bryan, S.A.

1995-07-01

The objective of this project is to determine the capacity for 137 Cs uptake by mixed metal ferrocyanides present in Hanford Site waste tanks, and to assess the potential for aggregation of these 137 Cs-exchanged materials to form ''hot-spots'' in the tanks. This research, performed at Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company, stems from concerns regarding possible localized radiolytic heating within the tanks. After ferrocyanide was added to 18 high-level waste tanks in the 1950s, some of the ferrocyanide tanks received considerable quantities of saltcake waste that was rich in 137 Cs. If radioactive cesium was exchanged and concentrated by the nickel ferrocyanide present in the tanks, the associated heating could cause tank temperatures to rise above the safety limits specified for the ferrocyanide-containing tanks, especially if the supernate in the tanks is pumped out and the waste becomes drier

Safety considerations to avoid current-induced skin burns in MRI

International Nuclear Information System (INIS)

Knopp, M.V.; Metzner, R.; Kaick, G. van; Brix, G.; Bundesamt fuer Strahlenschutz, Oberschleissheim

1998-01-01

The safety aspects of radiological methods continue to evolve. In this paper we report on two cases of skin burns in MRI caused by induced electrical current. A second- and a third-degree skin burn occurred during imaging in a 1.5 T system. The electromagnetic radiofrequency field inadvertently led to electrical currents caused by a conducting loop through the extremities and trunk. Skin burns induced by electrical current may occur in extremely rare cases even with standard MR imaging protocols operating within all current safety guidelines by inadvertently forming a closed conducting loop. By avoiding focal skin to skin contact of the extremities, this extremely rare adverse event can be avoided. (orig.) [de

Road safety development in Europe: a decade of changes (2001-2010).

Science.gov (United States)

Shen, Yongjun; Hermans, Elke; Bao, Qiong; Brijs, Tom; Wets, Geert

2013-11-01

To evaluate the road safety development of a country over time, the percentage change in the number of road fatalities is traditionally the main indicator. However, simply considering the reduction in the road fatalities may not correctly reflect the real improvement in road safety because the transport circumstances of a country underlying the road fatalities also change every year. In this study, we present a new way for measuring the road safety performance change over time, which is to use the technique of data envelopment analysis (DEA) and the Malmquist productivity index. In doing so, we can not only focus on the evolution of road safety final outcomes within a given period, but also take the changes of different measures of exposure in the same period into account. In the application, the DEA-based Malmquist productivity index (DEA-MI) is used to measure the extent to which the EU countries have improved their road safety performance over the period 2001-2010. More objective and insightful results are obtained compared to the ones based on the traditional indicator. The results show considerable road safety progress in most of the Member States during these ten years, and the fatality risk rather than the fatality number on Europe's roads has actually been reduced by approximately half. However, the situation differed considerably from country to country. The decomposition of the DEA-MI into 'efficiency change' and 'technical change' further reveals that the bulk of the improvement during the last decade was attained through the adoption of productivity-enhancing new technologies throughout the road transport sector in Europe, rather than through the relatively underperforming countries catching up with those best-performing ones. Copyright © 2013 Elsevier Ltd. All rights reserved.

77 FR 53164 - Railroad Workplace Safety; Adjacent-Track On-Track Safety for Roadway Workers

Science.gov (United States)

2012-08-31

...-0059, Notice No. 6] RIN 2130-AC37 Railroad Workplace Safety; Adjacent-Track On-Track Safety for Roadway... complex issues raised in both the petitions for reconsideration of the final rule published November 30... issues. One of the Petitions included a request for a delay in the effective date of the final rule until...

Safety balance: Analysis of safety systems; Bilans de surete: analyse par les organismes de surete

Energy Technology Data Exchange (ETDEWEB)

Delage, M; Giroux, C

1990-12-01

Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses.

Safety in nuclear power plants

International Nuclear Information System (INIS)

Koeberlein, K.

1987-01-01

In nuclear power plants large amounts of radioactive fission products ensue from the fission of uranium. In order to protect the environment, the radioactive material is confined in multiple 'activity barriers' (crystal matrix of the fuel, fuel cladding, coolant boundary, safety containment, reactor building). These barriers are protected by applying a defense-in-depth concept (high quality requirements, protection systems which recognize and terminate operational incidents, safety systems to cope with accidents). In spite of a favorable safety record of German nuclear power plants it is obvious - and became most evident by the Chernobyl accident - that absolute safety is not achievable. At Chernobyl, however, design disadvantages of that reactor type (like positive reactivity feedback of coolant voiding, missing safety containment) played an important role in accident initiation and progression. Such features of the Russian 'graphite-moderated pressure tube boiling water reactor' are different from those of light water reactors operating in western countries. The essential steps of the waste management of the nuclear fuel cycle ('Entsorgung') are the interim storage, the shipment, and the reprocessing of the spent fuel and the final repository of radioactive waste. Reprocessing means the separation of fossil material (uranium, plutonium) from radioactive waste. Legal requirements for radiological protection of the environment, which are identical for nuclear power plants and reprocessing plant, are complied with by means of comprehensive filter systems. Safety problems of a reprocessing plant are eased considerably by the fact that system pressures, process temperatures and energy densities are low. In order to confine the radioactive waste from the biosphere for a very long period of time, it is to be discarded after appropriate treatment into the deep geological underground of salt domes. (orig./HP) [de

Combined use of transcranial magnetic stimulation and metal electrode implants: a theoretical assessment of safety considerations

Science.gov (United States)

Golestanirad, Laleh; Rouhani, Hossein; Elahi, Behzad; Shahim, Kamal; Chen, Robert; Mosig, Juan R.; Pollo, Claudio; Graham, Simon J.

2012-12-01

This paper provides a theoretical assessment of the safety considerations encountered in the simultaneous use of transcranial magnetic stimulation (TMS) and neurological interventions involving implanted metallic electrodes, such as electrocorticography. Metal implants are subject to magnetic forces due to fast alternating magnetic fields produced by the TMS coil. The question of whether the mechanical movement of the implants leads to irreversible damage of brain tissue is addressed by an electromagnetic simulation which quantifies the magnitude of imposed magnetic forces. The assessment is followed by a careful mechanical analysis determining the maximum tolerable force which does not cause irreversible tissue damage. Results of this investigation provide useful information on the range of TMS stimulator output powers which can be safely used in patients having metallic implants. It is shown that conventional TMS applications can be considered safe when applied on patients with typical electrode implants as the induced stress in the brain tissue remains well below the limit of tissue damage.

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

Climate Considerations in Long-Term Safety Assessments for Nuclear Waste Repositories

Energy Technology Data Exchange (ETDEWEB)

Naeslund, Jens-Ove; Brandefelt, Jenny; Claesson Liljedahl, Lillemor [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)], E-mail: jens-ove.naslund@skb.se

2013-05-15

For a deep geological repository for spent nuclear fuel planned in Sweden, the safety assessment covers up to 1 million years. Climate scenarios range from high-end global warming for the coming 100 000 years, through deep permafrost, to large ice sheets during glacial conditions. In contrast, in an existing repository for short-lived waste the activity decays to low levels within a few tens of thousands of years. The shorter assessment period, 100 000 years, requires more focus on climate development over the coming tens of thousands of years, including the earliest possibility for permafrost growth and freezing of the engineered system. The handling of climate and climate change in safety assessments must be tailor-made for each repository concept and waste type. However, due to the uncertain future climate development on these vast time scales, all safety assessments for nuclear waste repositories require a range of possible climate scenarios.

49 CFR 1110.5 - Consideration of comments received.

Science.gov (United States)

2010-10-01

... 49 Transportation 8 2010-10-01 2010-10-01 false Consideration of comments received. 1110.5 Section 1110.5 Transportation Other Regulations Relating to Transportation (Continued) SURFACE TRANSPORTATION... PROCEEDINGS § 1110.5 Consideration of comments received. All timely comments will be considered before final...

Safety design

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Shiozawa, Shusaku

2004-01-01

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

NRC safety research in support of regulation, FY 1992

International Nuclear Information System (INIS)

1993-05-01

This report, the eighth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1992. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that a number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with the technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission's rules, regulatory guides, or other guidance

Safety in the final disposal of radioactive waste. Final report

International Nuclear Information System (INIS)

Broden, K.; Carugati, S.; Brodersen, K.

1997-12-01

During 1994-1997 a project on the disposal of radioactive waste was carried out as part of the NKS program. The objective of the project was to give authorities and waste producers in the Nordic countries background material for determinations about the management and disposal of radioactive waste. The project NKS/AFA-1 was divided into three sub-projects: AFA-1.1, AFA-1.2 and AFA-1.3. AFA-1.1 dealt with waste characterisation, AFA-1.2 dealt with performance assessment for repositories and AFA-1.3 dealt with Environmental Impact Assessment (EIA). The studies mainly focused on the management of long-lived low- and intermediate-level radioactive waste from research, hospitals and industry. The AFA-1.1 study included an overview on waste categories in the Nordic countries and methods to determine or estimate the waste content. The results from the AFA-1.2 study include a short overview of different waste management systems existing and planned in the Nordic countries. However, the main emphasis of the study was a general discussion of methodologies developed and employed for performance assessments of waste repositories. Some of the phenomena and interactions relevant for generic types of repository were discussed as well. Among the different approaches for the development of scenarios for safety and performance assessments one particular method, the Rock Engineering System (RES), was chosen to be tested by demonstration. The possible interactions and their safety significance were discussed, employing a simplified and generic Nordic repository system as the reference system. New regulations for the inventory of a repository may demand new assessments of old radioactive waste packages. The existing documentation of a waste package is then the primary information source although additional measurements may be necessary. (EG)

Safety in the final disposal of radioactive waste. Final report

Energy Technology Data Exchange (ETDEWEB)

Broden, K.; Carugati, S.; Brodersen, K. [and others

1997-12-01

During 1994-1997 a project on the disposal of radioactive waste was carried out as part of the NKS program. The objective of the project was to give authorities and waste producers in the Nordic countries background material for determinations about the management and disposal of radioactive waste. The project NKS/AFA-1 was divided into three sub-projects: AFA-1.1, AFA-1.2 and AFA-1.3. AFA-1.1 dealt with waste characterisation, AFA-1.2 dealt with performance assessment for repositories and AFA-1.3 dealt with Environmental Impact Assessment (EIA). The studies mainly focused on the management of long-lived low- and intermediate-level radioactive waste from research, hospitals and industry. The AFA-1.1 study included an overview on waste categories in the Nordic countries and methods to determine or estimate the waste content. The results from the AFA-1.2 study include a short overview of different waste management systems existing and planned in the Nordic countries. However, the main emphasis of the study was a general discussion of methodologies developed and employed for performance assessments of waste repositories. Some of the phenomena and interactions relevant for generic types of repository were discussed as well. Among the different approaches for the development of scenarios for safety and performance assessments one particular method, the Rock Engineering System (RES), was chosen to be tested by demonstration. The possible interactions and their safety significance were discussed, employing a simplified and generic Nordic repository system as the reference system. New regulations for the inventory of a repository may demand new assessments of old radioactive waste packages. The existing documentation of a waste package is then the primary information source although additional measurements may be necessary. (EG) 33 refs.

Climate considerations in long-term safety assessments for nuclear waste repositories.

Science.gov (United States)

Näslund, Jens-Ove; Brandefelt, Jenny; Liljedahl, Lillemor Claesson

2013-05-01

For a deep geological repository for spent nuclear fuel planned in Sweden, the safety assessment covers up to 1 million years. Climate scenarios range from high-end global warming for the coming 100 000 years, through deep permafrost, to large ice sheets during glacial conditions. In contrast, in an existing repository for short-lived waste the activity decays to low levels within a few tens of thousands of years. The shorter assessment period, 100 000 years, requires more focus on climate development over the coming tens of thousands of years, including the earliest possibility for permafrost growth and freezing of the engineered system. The handling of climate and climate change in safety assessments must be tailor-made for each repository concept and waste type. However, due to the uncertain future climate development on these vast time scales, all safety assessments for nuclear waste repositories require a range of possible climate scenarios.

Software and man-machine interface considerations for a nuclear plant computer replacement and upgrade project

International Nuclear Information System (INIS)

Diamond, G.; Robinson, E.

1984-01-01

Some of the key software functions and Man-Machine Interface considerations in a computer replacement and upgrade project for a nuclear power plant are described. The project involves the installation of two separate computer systems: an Emergency Response Facilities Computer System (ERFCS) and a Plant Process Computer System (PPCS). These systems employ state-of-the-art computer hardware and software. The ERFCS is a new system intended to provide enhanced functions to meet NRC post-TMI guidelines. The PPCS is intended to replace and upgrade an existing obsolete plant computer system. A general overview of the hardware and software aspects of the replacement and upgrade is presented. The work done to develop the upgraded Man-Machine Interface is described. For the ERFCS, a detailed discussion is presented of the work done to develop logic to evaluate the readiness and performance of safety systems and their supporting functions. The Man-Machine Interface considerations of reporting readiness and performance to the operator are discussed. Finally, the considerations involved in the implementation of this logic in real-time software are discussed.. For the PPCS, a detailed discussion is presented of some new features

Safety case plan 2008

International Nuclear Information System (INIS)

2008-07-01

Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy) Posiva is preparing to submit the construction license application for a spent fuel repository by the end of the year 2012. The long-term safety section supporting the license application is based on a safety case, which, according to the internationally adopted definition, is a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. In 2005, Posiva presented a plan to prepare such a safety case. The present report provides a revised plan of the safety case contents mentioned above. The update of the safety case plan takes into account the recommendations made by the Radiation and Nuclear Safety Authority (STUK) about improving the focus and further developing the plan. Accordingly, particular attention is given to the quality management of the safety case work, the management of uncertainties and the scenario methodology. The quality management is based on the ISO 9001:2000 standard process thinking enhanced with special features arising from STUK's YVL Guides. The safety case production process is divided into four main sub-processes. The conceptualisation and methodology sub-process defines the framework for the assessment. The critical data handling and modelling sub-process links Posiva's main technical and scientific activities to the production of the safety case. The assessment sub-process analyses the consequences of the evolution of the disposal system in various scenarios, classified either as part of the expected evolution or as disruptive scenarios. The compliance and confidence sub-process is responsible for final evaluation of compliance of the assessment results with the regulatory criteria and the overall confidence in the safety case. As in the previous safety case plan, the safety case will be based on several reports, but

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.

Technical bulletin : structural considerations for dam safety

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

This technical bulletin discussed issues related to the safety assessment of concrete water-retaining structures and timber dams. Structures reviewed in the paper included gravity dams; buttress dams; arch dams; spillway structures; intake structures; power plants; roller compacted concrete dams; and timber dams. A variety of issues related to the loss of cohesive bond and discontinuities in bedrock foundations were reviewed with reference to issues related to compressive strength, tensile strength, and shear strength. Static failure modes and failure mechanisms related to dam failures were also described. Visual indicators for potential failures include abutment and foundation movement, seepage, and structure movements. Loading combinations were discussed, and performance indicators for gravity dams were provided. Methods of analysis for considering load characteristics, structure types and geological conditions were also discussed. Modelling techniques for finite element analysis were also included. 16 refs., 3 tabs., 5 figs.

DOE high-level waste tank safety program Final report, Task 002

International Nuclear Information System (INIS)

1998-01-01

The overall objective of the work on Task 002 was to provide LANL with support to the DOE High-Level Waste Tank Safety program. The objective of the work was to develop safety documentation in support of the unsafe tank mitigation activities at Hanford. The work includes the development of safety assessment and an environmental assessment. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective are provided. The two tasks were: Task 2.1--safety assessment for instrumentation insertion; and Task 2.2--environmental assessment

Indicators to monitor NPP operational safety performance

International Nuclear Information System (INIS)

Gomez-Cobo, Ana

2002-01-01

Since December 1995 the IAEA activities on safety performance indicators focused on the elaboration of a framework for the establishment of an operational safety performance indicator programme. The development of this framework began with the consideration of the concept of NPP operational safety performance and the identification of operational safety attributes. For each operational safety attribute, overall indicators, envisioned as providing an overall evaluation of relevant aspects of safety performance, were established. Associated with each overall indicator is a level of strategic indicators intended to provide a bridge from overall to specific indicators. Finally each strategic indicator was supported by a set of specific indicators, which represent quantifiable measures of performance. The programme development was enhanced by pilot plant studies, conducted over a 15 month period from January 1998 to March 1999. The result of all this work is compiled in the IAEA-TECDOC-1141, to be published shortly. This paper presents a summary of this IAEA TECDOC. It describes the operational safety performance indicator framework proposed and discusses the results of and lessons learned from the pilot studies. Despite the efforts described, it is clear that additional research is still necessary in areas such as plant-specific adaptation of proposed frameworks in order to suit individual data collection systems and plant characteristics, indicator selection, indicator definition, goal setting, action thresholds, analysis of trends, indicator display systems, analysis of overall safety performance (i.e., aggregation or combination of indicators), safety culture indicators, qualitative indicators, and use of additional indicators to address issues such as industrial safety attitude and performance, staff welfare, and environmental compliance. This is the rationale for a new IAEA Coordinated Research Project on 'Development and application of indicators to monitor NPP

Nuclear safety policy statement in korea

International Nuclear Information System (INIS)

Kim, W.S.; Kim, H.J.; Choi, K.S.; Choi, Y.S.; Park, D.K.

2006-01-01

Full text: Wide varieties of programs to enhance nuclear safety have been established and implemented by the Korean government in accordance with the Nuclear Safety Policy Statement announced in September 1994. The policy statement was intended to set the long-term policy goals for 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 has been recognized as very effective in developing safety culture in nuclear-related organizations and also enhancing nuclear safety in Korea. However, ageing of operating nuclear power plants and increasing of new nuclear facilities have demanded a new comprehensive national safety policy to cover the coming decade, taking the implementation results of the policy statement of 1994 and the changing environment of nuclear industries into consideration. Therefore, the results of safety policy implementation have been reviewed and, considering changing environment and future prospects, a new nuclear safety policy statement as a highest level national policy has been developed. The implementation results of 11 regulatory policy directions such as the use of Probabilistic Safety Assessment, introduction of Periodic Safety Review, strengthening of safety research, introduction of Risk Based Regulation stipulated in the safety policy statement of 1994 were reviewed and measures taken after various symposia on nuclear safety held in Nuclear Safety Days since 1995 were evaluated. The changing international and domestic environment of nuclear industry were analysed and future prospects were explored. Based on the analysis and review results, a draft of new nuclear safety policy statement was developed. The draft was finalized after the review of many prominent experts in Korea. Considering changing environment and future prospects, new policy statement that will show government's persistent will for nuclear safety has been

21 CFR 25.16 - Public health and safety emergencies.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Public health and safety emergencies. 25.16... ENVIRONMENTAL IMPACT CONSIDERATIONS Agency Actions Requiring Environmental Consideration § 25.16 Public health... importance to the public health or safety, may make full adherence to the procedural provisions of NEPA and...

Considerations in the safety assessment of sealed nuclear facilities

International Nuclear Information System (INIS)

1991-06-01

This report is a part of the International Atomic Energy Agency's radioactive waste management programme, whose objective is to provide assistance to Member States in developing guidance for identifying safe alternatives for isolating radioactive waste from man and his environment. This report attempts to integrate information from the previous reports on decommissioning of nuclear facilities, mitigation of accidents at such facilities, and performance assessment of disposal systems to provide useful advice and qualitative guidance to those responsible for performance and safety assessments of sealed nuclear facilities by giving an overview of possible approaches and techniques for such assessments. In this context, the establishment of requirements and rules governing the radiological safety of personnel, the general public, and the environment for sealing and post-sealing activities will enable the choice of the most appropriated approach and help to promote consistency in both decommissioning and waste management standards. The near-field effects discussed in this document include gas generation, interactions of the groundwater and the residual water with other components of the system, thermal, thermo-mechanical, radiation effects and chemical and geochemical reactions. 59 refs, figs and tabs

Training Course for Compliance Safety and Health Officers. Final Report.

Science.gov (United States)

McKnight, A. James; And Others

The report describes revision of the Compliance Safety and Health Officers (CSHO) course for the Department of Labor, Occupational Safety and Health Administration (OSHA). The CSHO's job was analyzed in depth, in accord with OSHA standards, policies, and procedures. A listing of over 1,700 violations of OSHA standards was prepared, and specialists…

Safety and radiation protection aspects of the management of radioactive wastes of high level activity

International Nuclear Information System (INIS)

Candes, P.; Pradel, J.

1977-01-01

Appropriate consideration is given in France to safety and protection problems to be solved from the production up to the final disposal of radioactive wastes of high level activity. The first stage of the work consisted in emphasizing the various technical options. Different strategies appear to be possible, taking into account technical, political, and psychological difficulties. This results in evaluating the safety problems to be solved in the framework of those strategies. In this field, the main safety and protection principles do not differ from those which apply to other nuclear facilities. Nevertheless, duration is in most cases a quite different factor (thousands or millions of years). The question is then raised of evaluating the importance to be given to very remote consequences, both at philosophical and scientific levels. As a first result of those considerations, the application of the ''barrier'' concept is recommended. This concept is familiar to safety specialists. Different barriers for which particular problems are listed and evaluated, are defined. Another results with regard to radiation protection principles is to consider that if safety provisions should lead to a containment of radioactive products as efficient as possible, it would not be realistic to consider such a containment as absolute, in particular for disposal durations arising to thousands of years. It is therefore assumed that a limited radioactivity transfer should be taken into account, and its consequences for environment and man be calculated. This is especially true in the study of an appropriate site for final storage, and the study should necessarily include a detailed investigation of the retention characteristics of soil layers, and the implementation of appropriate models giving a sufficiently accurate evaluation of the consequences of transfers, including those related to the effect of various elements after their arrival into the biosphere. The authors review the

Nuclear Power Safety Reporting System. Final evaluation results

International Nuclear Information System (INIS)

Finlayson, F.C.; Newton, R.D.

1986-02-01

This document presents the results of a study conducted by the US Nuclear Regulatory Commission of an unobtrusive, voluntary, anonymous third-party managed, nonpunitive human factors data gathering system (the Nuclear power Safety Reporting System - NPSRS) for the nuclear electric power production industry. The data to be gathered by the NPSRS are intended for use in identifying and quantifying the factors that contribute to the occurrence of significant safety incidents involving humans in nuclear power plants. The NPSRS has been designed to encourage participation in the System through guarantees of reporter anonymity provided by a third-party organization that would be responsible for NPSRS management. As additional motivation to reporters for contributing data to the NPSRS, conditional waivers of NRC disciplinary action would be provided to individuals. These conditional waivers of immunity would apply to potential violations of NRC regulations that might be disclosed through reports submitted to the System about inadvertent, noncriminal incidents in nuclear plants. This document summarizes the overall results of the study of the NPSRS concept. In it, a functional description of the NPSRS is presented together with a review and assessment of potential problem areas that might be met if the System were implemented. Conclusions and recommendations resulting from the study are also presented. A companion volume (NUREG/CR-4133, Nuclear Power Safety Reporting System: Implementation and Operational Specifications'') presented in detail the elements, requirements, forms, and procedures for implementing and operating the System. 13 refs

Safety considerations and countermeasures against fire and explosion at an HTGR-hydrogen production system. Proposal of safety design concept

International Nuclear Information System (INIS)

Nishihara, T.; Hada, K.; Shibata, T.; Shiozawa, S.

1996-01-01

Establishment of safety design concept and countermeasures against fire and explosion accidents is among key safety-related issues in an HTGR-hydrogen production system. We propose the different safety design concepts depending upon the origin of fire and explosion which may happen in the HTGR-hydrogen production plant. Against fire and explosion originated outside the reactor building (R/B), namely in the area of hydrogen production plant, the safety design concept is primarily to take a safe distance for preventing the damage on safety-related items or a proof wall if necessary. Because the hydrogen production plant is designed in the same safety level as a conventional chemical plant. The safe distance is proposed to limit an incident overpressure to 10 kPa so as not to suffer any damage on the items and to limit a wall-averaged temperature of concrete structures of the R/B to 175degC according to the current regulation. On the other hand, against a potential possibility of explosion originated inside the R/B, the safety design concept is to minimize the possibility of explosion low enough to assume no occurrence inside the R/B. That is, the measure is to exclude a simultaneous failure of a secondary helium piping and an endothermic chemical reactor. Furthermore, in severe accident condition in which the explosion may be postulated a priori, an incidental overpressure of explosion inside the reactor containment vessel (C/V) should be limited so as not to fail the C/V through restricting the amount of combustible gas ingress into the C/V by means of a combination of C/V isolation valve installed in the helium piping and emergency shut off valve in the process feed gas line. (author)

Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Bourham, Mohamed A.; Gilligan, John G.

1999-08-14

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

Building Newcomer Competence for NPP Safety Assessment through Learning by Doing: Development of Level 1 Probabilistic Safety Assessment for Research Reactors

International Nuclear Information System (INIS)

Kuzmina, Irina

2014-01-01

Final remarks: • COMPASS-M project is a very fruitful study. 1. State-of-the-art competence for PSA technique in Malaysia (applicable to nuclear installations, incl. RR and NPP). 2. PSA model and report for the operating research reactor in Malaysia. → Risk estimate of core damage and ranking contributors to the risk; → Basis for further safety improvement of RR as appropriate. 3. Input for IAEA’s publications on PSA for research reactors. • The results will be available to interested Member States (security considerations be addressed); → Completion in mid-2014, paper to be published in PSAM-12; ► Managerial support is instrumental for success of learning-by-doing projects

The importance of safety in achieving the widespread use of hydrogen as a fuel

Energy Technology Data Exchange (ETDEWEB)

Edeskuty, F.J.

1997-09-01

The advantages of hydrogen fuel have been adequately demonstrated on numerous occasions. However, two major disadvantages have prevented any significant amount of corresponding development. These disadvantages have been in the economics of producing sufficient quantities of hydrogen and in the safety (both real and perceived) of its use. To date work has mostly been properly centered on solving the economic problems. However, a greater effort on the safety of new hydrogen systems now being proposed also deserves consideration. To achieve the greatest safety in the expansion of the use of hydrogen into its wide-spread use as a fuel, attention must be given to four considerations. These are, obtaining knowledge of all the physical principles involved in the new uses, having in place the regulations that allow the safe interfacing of the new systems, designing and constructing the new systems with safety in mind, and the training of the large number of people that will become the handlers of the hydrogen. Existing organizations that produce, transport, or use hydrogen on a large scale have an excellent safety record. This safety record comes as a consequence of dedicated attention to the above-mentioned principles. However, where these principles were not closely followed, accidents have resulted. Some examples can be cited. As the use of hydrogen becomes more widespread, there must be a mechanism for assuring the universal application of these principles. Larger and more numerous fleet operations with hydrogen fuel may be the best way to begin the indoctrination of the general public to the more general use of hydrogen fuel. Demonstrated safe operation with hydrogen is vital to its final acceptance as the fuel of choice.

Laboratory Safety Guide for Arkansas K-12 Schools.

Science.gov (United States)

Arkansas State Dept. of Education, Little Rock.

This document presents laboratory safety rules for Arkansas K-12 schools which were developed by the Arkansas Science Teachers Association (ASTA) and the Arkansas Department of Education (ADE). Contents include: (1) "Laboratory Safety Guide for Arkansas K-12 Schools"; (2) "Safety Considerations"; (3) "Safety Standards for Science Laboratories";…

Improving the regulation of safety at DOE nuclear facilities. Final report

International Nuclear Information System (INIS)

1995-12-01

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by DOE itself. The three major recommendations are: under any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE's nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

An index of financial safety of China

Directory of Open Access Journals (Sweden)

Xiaojun Jia

2015-04-01

Full Text Available Purpose: This paper combines a synthetic index system by the variables and evaluates China’s financial safety through the change of indexes in a comprehensive way. First of all, it builds the financial industry evaluation index system composed of 25indicators in terms of the operation of the financial industry and external economic environment and particularly takes into consideration factors which might trigger liquidity risks such as off-balance-sheet business, interbank business and shadow banking; then it selects 10 indicators to conduct empirical analysis and identifies the indicator weight through principal component analysis; finally it combines the financial safety indexes through the linear weighted comprehensive evaluation model.Design/methodology/approach: Synthesis of indexes is made by constructing a proper comprehensive evaluation mathematical model, integrating a number of evaluation indexes into one comprehensive evaluation index and then obtaining corresponding comprehensive evaluation results. In this paper, it selects 10 indexes to conduct empirical analysis and identifies the index weight through principal component analysis; finally it combines the financial safety indexes through the linear weighted comprehensive evaluation model. Principal component analysis (PCA is a statistical procedure that uses an orthogonal transformation to convert a set of observations of possibly correlated variables into a set of values of linearly uncorrelated variables called principal components. PCA was invented in 1901 and was later independently developed (and named by Harold Hotelling in the 1930s.Findings: From 2003 to 2013 China’s financial safety indexes fluctuated. From 2003 to 2007 indexes rose, which indicates China’s financial safety status gradually improved; from 2007 to 2009 indexes declined, which indicates due to the impact of subprime crisis, China’s financial safety status took a turn for the worse; from 2009 to 2012

System study of alternative waste management techniques: Final report

International Nuclear Information System (INIS)

1986-01-01

This report summarizes the important results achieved in conjunction with the Research and Development Priority ''Alternative Waste Management Techniques'' sponsored by the Federal Ministry of Research and Technology from 1981 to 1984. The subject of these studies was solely ''direct disposal'' of spent fuel elements. For this purpose a reference concept was selected from a variety of possible processes and engineered in detailed form by firms in the nuclear industry. Those who worked on the engineering concepts consider this waste management method technically feasible. Several disposal casks have been fabricated. The basic licensability of direct disposal can be evaluated on the basis of the documentation developed by the companies. The direct disposal method was compared with the ''integrated waste management concept'' using reference fuel cycles with respect to the following criteria: radiological safety and nuclear material safeguards and, in addition, economic and energy-policy aspects. It was found that with respect to radiological safety, including the long-term safety of the final repository, there are no significant differences between the two fuel cycles with and without reprocessing. With respect to the nuclear material safeguards of a final repository containing spent fuel elements, there are still a number of unanswered questions. From an economic standpoint, direct disposal will be more economical in the foreseeable future than integrated waste management. Quantification of the effects of one or the other waste management method on the national economy is not necessarily possible. Reprocessing is supported primarily by technological and energy-policy considerations. On the basis of the results, the conclusion is reached that reprocessing should be pursued further, but that at the same time direct disposal should be developed to the point of practical maturity

Management of radioactive material safety programs at medical facilities. Final report

International Nuclear Information System (INIS)

Camper, L.W.; Schlueter, J.; Woods, S.

1997-05-01

A Task Force, comprising eight US Nuclear Regulatory Commission and two Agreement State program staff members, developed the guidance contained in this report. This report describes a systematic approach for effectively managing radiation safety programs at medical facilities. This is accomplished by defining and emphasizing the roles of an institution's executive management, radiation safety committee, and radiation safety officer. Various aspects of program management are discussed and guidance is offered on selecting the radiation safety officer, determining adequate resources for the program, using such contractual services as consultants and service companies, conducting audits, and establishing the roles of authorized users and supervised individuals; NRC's reporting and notification requirements are discussed, and a general description is given of how NRC's licensing, inspection and enforcement programs work

76 FR 34773 - Final Safety Culture Policy Statement

Science.gov (United States)

2011-06-14

.... For example, industry representatives could begin to identify tacit organizational and personal goals... with NRC-regulated activities. Experience has shown that certain personal and organizational traits are present in a positive safety culture. A trait, in this case, is a pattern of thinking, feeling, and...

Development and validation of three-dimensional CFD techniques for reactor safety applications. Final report

International Nuclear Information System (INIS)

Buchholz, Sebastian; Palazzo, Simone; Papukchiev, Angel; Scheurer Martina

2016-12-01

The overall goal of the project RS 1506 ''Development and Validation of Three Dimensional CFD Methods for Reactor Safety Applications'' is the validation of Computational Fluid Dynamics (CFD) software for the simulation of three -dimensional thermo-hydraulic heat and fluid flow phenomena in nuclear reactors. For this purpose a wide spectrum of validation and test cases was selected covering fluid flow and heat transfer phenomena in the downcomer and in the core of pressurized water reactors. In addition, the coupling of the system code ATHLET with the CFD code ANSYS CFX was further developed and validated. The first choice were UPTF experiments where turbulent single- and two-phase flows were investigated in a 1:1 scaled model of a German KONVOI reactor. The scope of the CFD calculations covers thermal mixing and stratification including condensation in single- and two-phase flows. In the complex core region, the flow in a fuel assembly with spacer grid was simulated as defined in the OECD/NEA Benchmark MATIS-H. Good agreement are achieved when the geometrical and physical boundary conditions were reproduced as realistic as possible. This includes, in particular, the consideration of heat transfer to walls. The influence of wall modelling on CFD results was investigated on the TALL-3D T01 experiment. In this case, the dynamic three dimensional fluid flow and heat transfer phenomena were simulated in a Generation IV liquid metal cooled reactor. Concurrently to the validation work, the coupling of the system code ATHLET with the ANSYS CFX software was optimized and expanded for two-phase flows. Different coupling approaches were investigated, in order to overcome the large difference between CPU-time requirements of system and CFD codes. Finally, the coupled simulation system was validated by applying it to the simulation of the PSI double T-junction experiment, the LBE-flow in the MYRRA Spallation experiment and a demonstration test case simulating a pump trip

Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

International Nuclear Information System (INIS)

Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

2012-01-01

The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

Non-technical issues in safety assessments for nuclear disposal facilities

International Nuclear Information System (INIS)

Kallenbach-Herbert, Beate; Brohmann, Bettina

2010-09-01

The paper highlights that a comprehensive approach to safety affords the consideration of technology, organisation, personnel and social environment. In several safety relevant contexts of nuclear waste disposal these fields are closely interrelated. The approach for the consideration of socio-scientific aspects which is sketched in this paper supports the systematic treatment of safety relevant non-technical issues in the safety case or in safety assessments for a disposal project. Furthermore it may foster the dialogue among specialists from the technical, the natural- and the socio-scientific field on questions of disposal safety. In this way it may contribute to a better understanding among the affected scientific disciplines in nuclear waste disposal.

Management of radioactive material safety programs at medical facilities. Final report

Energy Technology Data Exchange (ETDEWEB)

Camper, L.W.; Schlueter, J.; Woods, S. [and others

1997-05-01

A Task Force, comprising eight US Nuclear Regulatory Commission and two Agreement State program staff members, developed the guidance contained in this report. This report describes a systematic approach for effectively managing radiation safety programs at medical facilities. This is accomplished by defining and emphasizing the roles of an institution`s executive management, radiation safety committee, and radiation safety officer. Various aspects of program management are discussed and guidance is offered on selecting the radiation safety officer, determining adequate resources for the program, using such contractual services as consultants and service companies, conducting audits, and establishing the roles of authorized users and supervised individuals; NRC`s reporting and notification requirements are discussed, and a general description is given of how NRC`s licensing, inspection and enforcement programs work.

Safety standards of IAEA for management systems

International Nuclear Information System (INIS)

Vincze, P.

2005-01-01

IAEA has developed a new series of safety standards which are assigned for constitution of the conditions and which give the instruction for setting up the management systems that integrate the aims of safety, health, life environment and quality. The new standard shall replace IAEA 50-C-Q - Requirements for security of the quality for safety in nuclear power plants and other nuclear facilities as well as 14 related safety instructions mentioned in the Safety series No. 50-C/SG-Q (1996). When developing of this complex, integrated set of requirements for management systems, the IAEA requirements 50-C-Q (1996) were taken into consideration as well as the publications developed within the International organisation for standardization (ISO) ISO 9001:2000 and ISO14001: 1996. The experience of European Union member states during the development, implementation and improvement of the management systems were also taken into consideration

Commercial-grade motors in safety-related applications: Final report

International Nuclear Information System (INIS)

Holzman, P.M.

1988-04-01

The objective of this project was to discuss the process necessary to utilize commercial grade equipment in safety related applications and to provide utilities with guidance for accepting commercial grade motors for safety-related applications. The generic commercial-grade concepts presented in this report can be successfully applied to motors. Commercial grade item utilization has the greatest applicability to motors in ''mild'' environments, because these motors are essentially similar to commercial grade motors in materials, construction methods, and capabilities. The acceptance process is less applicable to motors that are subject to ''harsh'' environments during postulated accidents, because of the unique design features and testing required to qualify these motors

Licensing procedures and safety criteria for research reactors in France

International Nuclear Information System (INIS)

Berry, J.L.; Lerouge, B.

1980-11-01

This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. Some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors. At the end, a few considerations are given to the consequences of the Osiris core conversion

16 CFR 1012.1 - General policy considerations; scope.

Science.gov (United States)

2010-01-01

... Policy, sets forth requirements for advance public notice, public attendance, and recordkeeping for... 16 Commercial Practices 2 2010-01-01 2010-01-01 false General policy considerations; scope. 1012.1 Section 1012.1 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION GENERAL MEETINGS POLICY-MEETINGS...

Safety Consideration for a Wet Interim Spent Fuel Store at Conceptual Design Stage

International Nuclear Information System (INIS)

Astoux, Marion

2014-01-01

EDF Energy plans to build and operate two UK EPRs at the Hinkley Point C (HPC) site in Somerset, England. Spent fuel from the UK EPRs will need to be managed from the time it is discharged from the reactor until it is ultimately disposed of and this will involve storing the spent fuel for a period in the fuel building and thereafter in a dedicated interim facility until it can be emplaced within the UK Geological Disposal Facility. EDF Energy has proposed that this interim store should be located on the Hinkley Point site which is consistent with UK policy. This Interim Spent Fuel Store (ISFS) will have the capability to store for at least one hundred years the spent fuel arising from the operation of the two EPR units (sixty years operation). Therefore, specificities regarding the lifetime of the facility have to be accounted for its design. The choice of interim storage technology was considered in some depth for the HPC project and wet storage (pool) was selected. The facility is currently at conceptual design stage, although its construction will be part of main site construction phase. Safety functions and safety requirements for this storage facility have been defined, in compliance with WENRA 'Waste and Spent Fuel Storage - Safety Reference Level Report' and IAEA Specific Safety Guide no. 15 'Storage of Spent Nuclear Fuel'. EDF technical know-how, operational feedback on existing storage pools, UK regulatory context and Fukushima experience feedback have also been accounted for. Achievement of the safety functions as passively as reasonably practicable is a key issue for the design, especially in accident situations. Regarding lifetime aspects, ageing management of equipments, optimisation of the refurbishment, climate change, passivity of the facility, and long-term achievement of the safety functions are among the subjects to consider. Adequate Operational Limits and Conditions will also have to be defined, to enable the long-term achievement of the safety

HSE's safety assessment principles for criticality safety

International Nuclear Information System (INIS)

Simister, D N; Finnerty, M D; Warburton, S J; Thomas, E A; Macphail, M R

2008-01-01

The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf). (memorandum)

Safety analysis procedures for PHWR

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

2004-03-01

The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

A study on LAN applications in nuclear safety systems

International Nuclear Information System (INIS)

Kim, Sung; Lee, Young Ryul; Koo, Jun Mo; Han, Jai Bok

1995-01-01

It is a general tendency to digitalize the conventional relay based I and C systems in nuclear power plant. But, the digitalisation of nuclear safety systems has many a difficulty to surmount. The typical one thing of many difficulties is the data communication problem between local controllers and systems. The network architecture built with LAN (Local Area Network) in digital systems of the other industries are general. But in case of nuclear safety systems many considerations in point of safety and license are required to implement it in the field. In this parer, some considerations for applying LAN in nuclear safety systems were reviewed

NMC and A and nuclear criticality safety systems integration: A prospective way for enhancement of the nuclear industry facilities safety

International Nuclear Information System (INIS)

Ryazanov, Boris G.; Sviridov, Victor I.; Frolov, Vladimir V.; Shvedov, Maxim O.; Mclaughlin, Thomas P.; Pruvost, Norman L.

2003-01-01

A considerable body of data has now been acquired about the principles, parameters and consequences of nuclear (criticality) accidents at facilities of the atomic industry in Russia, the United States, Great Britain and Japan. The total number of such accidents stands at 22. Russian and US specialists have prepared a rather extensive survey and analysis of these accidents. The final and important section of this survey is the lessons implied by the results of analysis of these 22 accidents. Among these lessons is the necessity of unconditional enforcement of control over the movement and transformations of special nuclear materials (SNM), and in particular fissile materials, (those SNMs with criticality accident concerns) during production and processing. Inadequacies in such control have been among the causes of most of the accidents that have occurred. Nuclear materials control and accounting (MC and A) for the purpose of ensuring storage reliability and nonproliferation safeguards is a major task of nuclear facilities in any nation. MC and A systems use the latest techniques and hardware for periodic control of SNM in specifically organized material balance areas. Immediate checking, periodic inventory of SNM, and measurements of the parameters of SNM at key points are the main sources of data for these systems. Data about the presence and sites of location of SNM in material balance areas that are acquired in inventories can be used for objective assessment of the status of nuclear safety. On the other hand, the inventory itself involves performance of operations that are unlike routine process engineering, and require special consideration of nuclear safety. Use of the techniques and hardware of MC and A systems not only for purposes of storage reliability, but also to ensure nuclear safety, will reduce the risk of nuclear accidents. This paper gives a concise overview of nuclear accidents that have occurred due to inadequacies in MC and A, and demonstrates

NRC safety research in support of regulation. Volume 8, FY 1993

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1993. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with sound technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission`s rules, regulatory guides, or other guidance.

Safety factors for neutron fluences in NPP safety assessment

International Nuclear Information System (INIS)

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

2016-01-01

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

Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

International Nuclear Information System (INIS)

Bourham, Mohamed A.; Gilligan, John G.

1999-01-01

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

Effects of Economic Recession on Road Safety Indexes

Energy Technology Data Exchange (ETDEWEB)

Rojo, M.; Gonzalo-Orden, H.; Linares, A.; Olio, L. dell’

2016-07-01

During the last years, the investment in both construction and conservation of transportinfrastructures has been considerably reduced in several countries, as Spain. After anumber of years in which economic circumstances have forced Governments to reducebudgets earmarked for the maintenance and creation of new ways, it is interesting toanalyze whether this has taken a toll on accident rates.The paper evaluates if there are significant changes in the road safety through these yearsin Spain, comparing the annual statistics concerning investment in infrastructure andaccidents. Thus, the classical risk, mortality and severity indexes have been analyzed tounderstand their real trends. Finally, through linear regression techniques, it is shown howthese trends are related to the budgets invested each year, in order to draw interestingconclusions about the effect of their reduction. (Author)

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

International Nuclear Information System (INIS)

Lee, J.H.; Oka, Y.; Koshizuka, S.

1999-01-01

The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics

Statistical considerations of graphite strength for assessing design allowable stresses

International Nuclear Information System (INIS)

Ishihara, M.; Mogi, H.; Ioka, I.; Arai, T.; Oku, T.

1987-01-01

Several aspects of statistics need to be considered to determine design allowable stresses for graphite structures. These include: 1) Statistical variation of graphite material strength. 2) Uncertainty of calculated stress. 3) Reliability (survival probability) required from operational and safety performance of graphite structures. This paper deals with some statistical considerations of structural graphite for assessing design allowable stress. Firstly, probability distribution functions of tensile and compressive strengths are investigated on experimental Very High Temperature candidated graphites. Normal, logarithmic normal and Weibull distribution functions are compared in terms of coefficient of correlation to measured strength data. This leads to the adaptation of normal distribution function. Then, the relation between factor of safety and fracture probability is discussed on the following items: 1) As the graphite strength is more variable than metalic material's strength, the effect of strength variation to the fracture probability is evaluated. 2) Fracture probability depending on survival probability of 99 ∼ 99.9 (%) with confidence level of 90 ∼ 95 (%) is discussed. 3) As the material properties used in the design analysis are usually the mean values of their variation, the additional effect of these variations on the fracture probability is discussed. Finally, the way to assure the minimum ultimate strength with required survival probability with confidence level is discussed in view of statistical treatment of the strength data from varying sample numbers in a material acceptance test. (author)

Chemical considerations in severe accident analysis

International Nuclear Information System (INIS)

Malinauskas, A.P.; Kress, T.S.

1988-01-01

The Reactor Safety Study presented the first systematic attempt to include fission product physicochemical effects in the determination of expected consequences of hypothetical nuclear reactor power plant accidents. At the time, however, the data base was sparse, and the treatment of fission product behavior was not entirely consistent or accurate. Considerable research has since been performed to identify and understand chemical phenomena that can occur in the course of a nuclear reactor accident, and how these phenomena affect fission product behavior. In this report, the current status of our understanding of the chemistry of fission products in severe core damage accidents is summarized and contrasted with that of the Reactor Safety Study

Food Safety and Sanitary Practices of Selected Hotels in Batangas Province, Philippines: Basis of Proposed Enhancement Measures

Directory of Open Access Journals (Sweden)

April M. Perez

2017-02-01

Full Text Available This study assessed the extent of food safety and sanitary practices of selected hotels in Batangas province as basis of proposed enhancement measures. The study utilized descriptive method to describe food safety and sanitary practices of selected hotels in Batangas province with a total of 8 hotels (256 respondents. Purposive sampling was used in the study. The questionnaires were designed using the provision of the Sanitation Code of the Philippines, validated and finalized to come up with legitimate results. The study showed that there were eight (8 hotel respondents classified as two, three, four star with considerable years of experience and adequate number of employees. The hotels demonstrated the food safety and sanitary practices always in the areas of restaurant, bar service, catering and banquet and room service. The significant pair-wise comparison for restaurant, bar service, catering and banquet and room service shows that 2 star hotels greatly differs. The researcher recommends that the management should maintain high standard of food safety and sanitary practices among its staff, upgrade the food safety and sanitary practices for food safety accreditation, continuous training of the hotel managers/employees on food safety and sanitary practices.

Application of the new requirements of safety of the IAEA for the previous management to the final disposal of radioactive waste in the region: a personal vision

International Nuclear Information System (INIS)

Sed, Luis Andres Jova

2013-01-01

The work includes the requirements for the responsibilities associated with the management prior to the final disposal of radioactive waste and as they are referred to in the Region. Also discusses the requirements for the main stages of the management prior to the final disposal of radioactive waste. A very important section of the new requirements is that establish requirements for safe operation of facilities management prior to the final disposal of radioactive wastes and the implementation of activities under conditions of safety and development. The work is emphatic on the importance of safety justification since the beginning of the development of a facility as a basis for the decision-making and approval process. Emphasis is also on the gradual approach which should provide for the collection, analysis and interpretation of the relevant technical data, plans for the design and operation, and the formulation of the justification of the security. This paper gives a personal view of the situation in the Region

Food-packaging interactions influencing quality and safety.

Science.gov (United States)

Hotchkiss, J H

1997-01-01

Interactions between foods and packaging can be detrimental to quality and/or safety. Changes in product flavour due to aroma sorption and the transfer of undesirable flavours from packaging to foods are important mechanisms of deterioration when foods are packaged in polymer-based materials. Careful consideration must be given to those factors affecting such interactions when selecting packaging materials in order to maximize product quality, safety, and shelf-life while minimizing undesirable changes. Product considerations include sensitivity to flavour and related deteriorations, colour changes, vitamin loss, microbial activity, and amount of flavour available. Storage considerations include temperature, time, and processing method. Polymer considerations include type of polymer and processing method, volume or mass of polymer to product ratio, and whether the interaction is Fickian or non-Fickian. Methodology to determine the extent of such interactions must be developed. Direct interactions between food and packaging are not necessarily detrimental. The same principles governing undesirable interactions can be used to affect desirable outcomes. Examples include films which directly intercept or absorb oxygen, inhibit microorganisms, remove undesirable flavours by sorption, or indicate safety and product shelf-life.

The importance of safety in achieving the widespread use of hydrogen as a fuel

Energy Technology Data Exchange (ETDEWEB)

Edeskuty, F.J. [Los Alamos National Laboratory, Los Alamos, NM (United States)

1998-07-01

The advantages of hydrogen fuel have been adequately demonstrated on numerous occasions. However, two major disadvantages have prevented any significant amount of corresponding development. These disadvantages have been in the economics of producing sufficient quantities of hydrogen and in the safety (both real and perceived) of its use. To date, work has mostly been properly centered on solving the economic problems. However, a greater effort on the safety of new hydrogen systems now being proposed also deserves consideration. To achieve the greatest safety in the expansion of the use of hydrogen into its wide-spread use as a fuel, attention must be given to four considerations. These are, obtaining knowledge of all the physical principles involved in the new uses, using that knowledge to put in place the regulations that allow the safe interfacing of the new systems, designing and constructing the new systems with safety in mind, and the training of the large number of people that will become the handlers of the hydrogen. Existing organizations that produce, transport, or use hydrogen on a large scale have an excellent safety record. This safety record comes as a consequence of dedicated attention to the above-mentioned principle. However, where these principles were not closely followed, accidents have resulted. Some examples are cited. As the use of hydrogen becomes more wide-spread, there must be a mechanism for assuring the universal application of these principles. Larger and more numerous fleet operations with hydrogen fuel may be the best way to begin the indoctrination of the general public to the more general use of hydrogen fuel. Demonstrated safe operation with hydrogen is vital to its final acceptance as the fuel of choice. 18 refs.

What Are the Safety Considerations for Insulin Control for Athletes?

Science.gov (United States)

McDaniel, Larry W.; Olson, Sara; Gaudet, Laura; Jackson, Allen

2010-01-01

Athletes diagnosed with diabetes may have difficulty with their blood sugar levels fluctuating during intense exercise. Considerations for athletes with insulin concerns may range anywhere from exercise rehabilitation to the use of an automatic insulin pump. The automatic insulin pump is a small battery-operated device about the size of a pager.…

Initial integration of accident safety, waste management, recycling, effluent, and maintenance considerations for low-activation materials

International Nuclear Information System (INIS)

Piet, S.J.; Herring, J.S.; Cheng, E.T.; Fetter, S.

1991-01-01

A true low-activation material should ideally achieve all of the following objectives: 1. The possible prompt dose at the site boundary from 100% release of the inventory should be <2 Sv (200 rem); hence, the design would be inherently safe in that no possible accident could result in prompt radiation fatalities. 2. The possible cancers from realistic releases should be limited such that the accident risk is <0.1%/yr of the existing background cancer risk to local residents. This includes consideration of elemental volatility. 3. The decay heat should be limited so that active mitigative measures are not needed to protect the investment from cooling transients; hence, the design would be passively safe with respect to decay heat. 4. Used materials could be either recycled or disposed of as near- surface waste. 5. Hands-on maintenance should be possible around coolant system piping and components such as the heat exchanger. 6. Effluent of activation products should be minor compared to the major challenge of limiting tritium effluents. The most recent studies in these areas are used to determine which individual elements and engineering materials are low activation. Grades from A (best) to G (worst) are given to each element in the areas of accident safety, recycling, and waste management. Structure/fluid combinations are examined for low-activation effluents and out-of-blanket maintenance. The lowest activation structural materials are silicon carbide, vanadium alloys, and ferritic steels. Impurities and minor alloying constituents must be carefully considered. The lowest activation coolants are helium, water, FLiBe, and lithium. The lowest activation breeders are lithium, lithium oxide, lithium silicate, and FLiBe. Designs focusing on these truly low-activation materials will help achieve the excellent safety and environmental potential of fusion energy

Rankine bottoming cycle safety analysis. Final report

Energy Technology Data Exchange (ETDEWEB)

Lewandowski, G.A.

1980-02-01

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

Pharmacy technician-to-pharmacist ratios: a state-driven safety and quality decision.

Science.gov (United States)

Bess, D Todd; Carter, Jason; DeLoach, Lindsey; White, Carol L

2014-01-01

To discuss the policy of pharmacy technician-to-pharmacist ratios by comparing Florida as an example of legislative-led authority versus Tennessee as an example of board of pharmacy-led ruling. Over the past 2 years, the Florida legislature has debated the issue of pharmacy staffing ratios, initially leaving the Florida Board of Pharmacy with little authority to advocate for and enact safe technician staffing ratios. Anticipating this situation, the Tennessee Board of Pharmacy created rules to meet pharmacy staffing needs while protecting the authority of the pharmacist-in-charge and promoting patient safety. Before enacting rules, members of the board toured the state and talked about proposed rule changes with pharmacists. The final rule sets the pharmacy technician-to-pharmacist ratio at 2:1 but permits a 4:1 ratio based on public safety considerations and availability of at least two Certified Pharmacy Technicians. Pharmacists and leaders within the profession should conduct further research on appropriate and safe ratios of pharmacy technicians to pharmacists, with a focus on safety and quality of care.

Safety in waste management plants: An Indian perspective

International Nuclear Information System (INIS)

Shekhar, P.; Ozarde, P.D.; Gandhi, P.M.

2000-01-01

Assurance of safety of public and plant workers and protection of the environment are prime objectives in the design and construction of Waste Management Plants. In India, waste management principles and strategies have been evolved in accordance with national and international regulations and standards for radiation protection. The regulations governing radiation protection have a far-reaching impact on the management of the radioactive waste. The wastes arise at each stages of the fuel cycle with varying chemical nature, generation rate and specific activity levels depending upon the type of the facility. Segregation of waste based on its chemical nature and specific activity levels is an essential feature, as its aids in selection of treatment and conditioning process. Selection of the process, equipment and materials in the plant, are governed by safety consideration alongside factors like efficiency and simplicity. The plant design considerations like physical separation, general arrangement, ventilation zoning, access control, remote handling, process piping routing, decontamination etc. have major role in realizing waste safety. Stringent quality control measures during all stages of construction have helped in achieving the design intended safety. These aspects together with operating experience gained form basis for the improved safety features in the design and construction of waste management plants. The comprehensive safety is derived from adoption of waste management strategies and appropriate plant design considerations. The paper briefly brings safety in waste management programme in India, in its current perspective. (author)

The use of probabilistic safety assessments for improving nuclear safety in Europe

International Nuclear Information System (INIS)

Birkhofer, A.

1992-01-01

The political changes in Europe broadened the scope of international nuclear safety matters considerably. The Western world started to receive reliable and increasingly detailed information on Eastern European nuclear technology and took note of a broad range of technical and administrative problems relevant for nuclear safety in these countries. Reunification made Germany a focus of information exchange on these matters. Here, cooperation with the former German Democratic Republic and with other Eastern European countries as well as safety analyses of Soviet-built nuclear power plants started rather early. Meanwhile, these activities are progressing toward all-European cooperation in the nuclear safety sector. This cooperation includes the use of probabilistic safety assessments (PSAs) addressing applications in both Western and Eastern Europe as well as the further development of this methodology in a converging Europe

Safety requirements to be met in final storage of heat-producing waste an evaluation of the BMU draft

International Nuclear Information System (INIS)

Thomauske, B.

2008-01-01

The German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) on August 12, 2008 published a July 29, 2008 draft of the ''Safety Requirements to Be Met in Final Storage of Heat-producing Radioactive Waste.'' As announced by the BMU, these safety requirements are to bring up to the state of the art the safety criteria of 1983. Over a couple of years, efforts had been made to adapt the criteria to the internationally accepted standard as demanded by the Advisory Committees on Reactor Safeguards (RSK) and Radiation Protection (SSK). The main changes made by the BMU are the introduction of a phased procedure in building repositories. A phased plans approval procedure under the Atomic Energy Act has been foreseen by the Ministry for this purpose. In addition, the draft provides for the introduction of a risk-based goal of protection. To ensure retrievability of the waste, the casks are to have a demonstrated service life of 500 years. The BMU draft safety requirements are unable to bring the safety criteria of 1983 up to the current state of the art. Here are the key points of criticism: - A risk-based goal of protection is introduced. The yardstick to be applied is to be defined in a guideline yet to be elaborated. As a consequence, the draft lacks substance. - As in licensing of nuclear facilities, the licensing procedure provides for a phased plans approval procedure for exploration. This analogy does not exist, as exploration is not the first phase of the plant to be built but a measure which is a precondition for obtaining a permit for construction and operation. - The information contained in the draft indicates that, contrary to international recommendations, it tightens the goal of protection by more than one order of magnitude. - The requirements to be met by the casks because of retrievability impose constraints on solutions optimized for safety in emplacement technology. - The risk-based approach is not mature and is

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

In 1995 Mr. Joseph DiNunno of the Defense Nuclear Facilities Safety Board issued an approach to describe the concept of an integrated safety management program which incorporates hazard and safety analysis to address a multitude of hazards affecting the public, worker, property, and the environment. Since then the U S . Department of Energy (DOE) has adopted a policy to systematically integrate safety into management and work practices at all levels so that missions can be completed while protecting the public, worker, and the environment. While the DOE and its contractors possessed a variety of processes for analyzing fire hazards at a facility, activity, and job; the outcome and assumptions of these processes have not always been consistent for similar types of hazards within the safety analysis and the fire hazard analysis. Although the safety analysis and the fire hazard analysis are driven by different DOE Orders and requirements, these analyses should not be entirely independent and their preparation should be integrated to ensure consistency of assumptions, consequences, design considerations, and other controls. Under the DOE policy to implement an integrated safety management system, identification of hazards must be evaluated and agreed upon to ensure that the public. the workers. and the environment are protected from adverse consequences. The DOE program and contractor management need a uniform, up-to-date reference with which to plan. budget, and manage nuclear programs. It is crucial that DOE understand the hazards and risks necessarily to authorize the work needed to be performed. If integrated safety management is not incorporated into the preparation of the safety analysis and the fire hazard analysis, inconsistencies between assumptions, consequences, design considerations, and controls may occur that affect safety. Furthermore, confusion created by inconsistencies may occur in the DOE process to grant authorization of the work. In accordance with

Investigation of nuclear power safety objects

International Nuclear Information System (INIS)

2003-09-01

It is a report of ground and concept of nuclear safety objects and future issues in Japan, which has investigated by the Committee of Experts on Investigation of Nuclear Safety Objects in the Nuclear Safety Research Association. The report consisted of member of committee, main conclusions and five chapters. The first chapter contains construction of safety objects and range of object, the second chapter qualitative safety objects, the third chapter quantitative safety objects, the forth subsiding objects and the fifth other items under consideration. The qualitative safety objects on individual and society, the quantitative one on effects on health and social cost, aspect of safety objects, relation between radiation protection and safety objects, practical objective values and earthquake are stated. (S.Y.)

Some considerations for assurance of reactor safety from experiences in research reactors

International Nuclear Information System (INIS)

Okamoto, Sunao; Nishihara, Hideaki; Shibata, Toshikazu

1981-01-01

For the purpose of assuring reactor safety and strengthening research in the related fields, a multi-disciplinary group was formed among university researchers, including social scientists, with a special allocation of the Grant-in-Aid from the Ministry of Education, Science and Culture. An excerpt from the first year's report (1979 -- 1980) is edited here, which contains an interpretation of Murphy's reliability engineering law, a scope of reactor diagnostic studies to be pursued at universities, and safety measures already implemented or suggested to be implemented in university research reactors. (author)

Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices

International Nuclear Information System (INIS)

1995-12-01

The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE's nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation

Objectives of safety evaluation

International Nuclear Information System (INIS)

Rosen, M.

1980-01-01

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

Medical safety in boxing: administrative, ethical, legislative, and legal considerations.

Science.gov (United States)

Schwartz, Michael B

2009-10-01

The roles and responsibilities of the ringside physician are complex and have evolved into a unique specialty in sport medicine. In addition to the medical aspects of ringside medicine, the doctor is now responsible for many administrative, ethical, and legal considerations. This article reviews and details the numerous roles the ringside physician plays in the sport of boxing.

Occupational health and safety considerations for women employed in core mining positions

Directory of Open Access Journals (Sweden)

Doret Botha

2015-06-01

Full Text Available Orientation: Despite various liberalisation and feminisation processes with regard to gender and sex roles, traditionalistic typologies, especially in terms of occupational roles, are seemingly very reluctant to disappear from relevant theoretical discourses, as well as in practice. One of the main issues remains the terrain of physical work. Although women all over the world have been involved in mining activities for centuries, the mining industry has not been an obvious career choice for women. In South Africa, new mining legislation aims to rectify previous inequalities and disadvantages in the mining sector and specifically provides for the inclusion of women in core mining activities. Although well intended, women’s involvement in the core business of mining also exposes them to the various hazards related to mine work. Research purpose: This research determined perceptions regarding the health and safety of women working in core mining positions. Motivation for the study: Currently there is a paucity of published data regarding health and safety challenges pertaining to women employed in the core business of mining. Method: Quantitative and qualitative research paradigms were used (mixed method research design. Quantitative data were collected by means of a structured questionnaire. Qualitative data were collected by means of individual interviews and group interviews. Main findings: From the literature review and the empirical findings it is evident that various factors (physical work capacity, anthropometry and body composition, personal protective equipment, treatment during pregnancy and security measures need to be considered to ensure the health and safety of women employed in core mining positions. Practical/managerial implications: It is evident from the research that exceptional attention should be given to the promotion of the health and safety of women working in the core business of mines to sustain their involvement in the

Safety-evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 16

International Nuclear Information System (INIS)

1983-08-01

This supplement addresses certain allegations received by the staff concerning safety deficiencies at the Diablo Canyon project, specifically those relative to design aspects of the component cooling water system (CCWS), including its seismic classification, single failure considerations, and its heat removal capabilities. It is based on the staff's re-review of the Diablo Canyon Final Safety Analysis Report (FSAR); on a site walkdown conducted on January 12, 1983; on information obtained during meetings with the applicant dated March 15, 18, and 25, April 4, 7, and 15, and May 3 and 18, 1983

Safety case for license application for a final repository: The French example

International Nuclear Information System (INIS)

Boissier, Fabrice; Voinis, Sylvie

2014-01-01

The reversible repository in a deep geological formation is the French reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste (HLW and ILW). Twenty years of R and D work and conceptual and basic studies since the first French Act of 1991 led, in particular, to a feasibility demonstration in 2005. According to the French Act on Radioactive Waste of 28 of June 2006, Andra shall design a reversible repository in order to apply for license in 2015. In response to this demand, Andra developed the industrial project known as 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. Two years before applying for authorisation, Andra's project is now focusing on three main targets: developing Cigeo's industrial design, preparing the authorisation process through increased exchanges with stakeholders and the preparation of a safety case to support authorisation application. The latter draws on the previous safety cases of 2005 and 2009, which give a sound basis to assess Cigeo's safety, both for the operational and post-closure periods. In this new stage of the project, the challenging issues for the preparation of the safety case are the following: - to identify the various regulatory frameworks (nuclear and non-nuclear) and guides applicable to the facility; - to ensure that the industrial design complies in particular with the safety requirements as presented in the safety case and its supporting safety assessment; - to identify crucial inputs (R and D, tests,...) needed to support the authorisation application, in particular, to bring convincing arguments to assess the technical feasibility of the design and when appropriate its ability to meet the safety requirements; - to ensure that all the requirements from previous regulatory and peer reviews (national and international?) are taken into account. (authors)

Safety and performance assessment of geologic disposal systems for nuclear wastes

International Nuclear Information System (INIS)

Peltonen, E.

1987-01-01

This thesis presents a methodology for the safety and performance assesment of final disposal of nuclear wastes into crystalline bedrock. The applicability of radiation protection objectives is discussed, as well as the goals of the assessment in the various repository system development phases. Due consideration is given to the description of the pertinent analysis methods and to the comprehensive model system. The methodology has been applied to assess the acceptability of the basic disposal concepts and to study the possibilities for the optimization of protection. Furthermore, performance of different components in the multiple barrier disposal systems is estimated. The waste types dealt with are low- and intermediate-level waste as well as high-level spent nuclear fuel from a nuclear power plant. In addition, an option of high-level vitrified waste from reprocessing of spent fuel is taken into account. On the basis of the various analyses carried out it can be concluded that the disposal of different nuclear wastes in the Finnish bedrock in properly designed repositories meets the radiation protection objectives with good confidence. In addition, the studies indicate that the safety margins are considerable. This is due to the fact that the overall performance of the multiple barrier disposal systems analysed is not sensitive to possible unfavourable changes in barrier properties. From the optimization of protection point of view it can be concluded that there is no need to develop more effective repository designs than those analysed in this thesis. In fact, the results indicate that the most sophisticated designs have already gone beyond an optimal level of safety

A Study on the Allowable Safety Factor of Cut-Slopes for Nuclear Facilities

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Soo; Yee, Eric [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

In this study, the issues of allowable safety factor design criteria for cut-slopes in nuclear facilities is derived through case analysis, a proposed construction work slope design criteria that provides relatively detailed conditions can be applied in case of the dry season and some unclear parts of slope design criteria be modified in case of the rainy season. This safety factor can be further subdivided into two; normal and earthquake factors, a factor of 1.5 is applied for normal conditions and a factor of 1.2 is applied for seismic conditions. This safety factor takes into consideration the effect of ground water and rainfall conditions. However, no criteria for the case of cut-slope in nuclear facilities and its response to seismic conditions is clearly defined, this can cause uncertainty in design. Therefore, this paper investigates the allowable safety factor for cut-slopes in nuclear facilities, reviews conditions of both local and international cut-slope models and finally suggests an alternative method of analysis. It is expected that the new design criteria adequately ensures the stability of the cut-slope to reflect clear conditions for both the supervising and design engineers.

Safety Supervisory Strategy for an Upper-Limb Rehabilitation Robot Based on Impedance Control

Directory of Open Access Journals (Sweden)

Lizheng Pan

2013-02-01

Full Text Available User security is an important consideration for robots that interact with humans, especially for upper-limb rehabilitation robots, during the use of which stroke patients are often more susceptible to injury. In this paper, a novel safety supervisory control method incorporating fuzzy logic is proposed so as to guarantee the impaired limb's safety should an emergency situation occur and the robustness of the upper-limb rehabilitation robot control system. Firstly, a safety supervisory fuzzy controller (SSFC was designed based on the impaired-limb's real-time physical state by extracting and recognizing the impaired-limb's tracking movement features. Then, the proposed SSFC was used to automatically regulate the desired force either to account for reasonable disturbance resulting from pose or position changes or to respond in adequate time to an emergency based on an evaluation of the impaired-limb's physical condition. Finally, a position-based impedance controller was implemented to achieve compliance between the robotic end-effector and the impaired limb during the robot-assisted rehabilitation training. The experimental results show the effectiveness and potential of the proposed method for achieving safety and robustness for the rehabilitation robot.

Safety Aspects in Radioactive Waste Management

Directory of Open Access Journals (Sweden)

Peter W. Brennecke

2007-01-01

Full Text Available In recent years, within the framework of national as well as international programmes, notable advances and considerable experience have been reached, particularly in minimising of the production of radioactive wastes, conditioning and disposal of short-lived, low and intermediate level waste, vitrification of fission product solutions on an industrial scale and engineered storage of long-lived high level wastes, i.e. vitrified waste and spent nuclear fuel. Based on such results, near-surface repositories have successfully been operated in many countries. In contrast to that, the disposal of high level radioactive waste is still a scientific and technical challenge in many countries using the nuclear power for the electricity generation. Siting, planning and construction of repositories for the high level wastes in geological formations are gradually advancing. The site selection, the evaluation of feasible sites as well as the development of safety cases and performance of site-specific safety assessments are essential in preparing the realization of such a repository. In addition to the scientific-technical areas, issues regarding economical, environmental, ethical and political aspects have been considered increasingly during the last years. Taking differences in the national approaches, practices and the constraints into account, it is to be recognised that future developments and decisions will have to be extended in order to include further important aspects and, finally, to enhance the acceptance and confidence in the safety-related planning work as well as in the proposed radioactive waste management and disposal solutions.

Radiation safety considerations and compliance within equine veterinary clinics: Results of an Australian survey

International Nuclear Information System (INIS)

Surjan, Y.; Ostwald, P.; Milross, C.; Warren-Forward, H.

2015-01-01

Objective: To examine current knowledge and the level of compliance of radiation safety principles in equine veterinary clinics within Australia. Method: Surveys were sent to equine veterinary surgeons working in Australia. The survey was delivered both online and in hardcopy format; it comprised 49 questions, 15 of these directly related to radiation safety. The participants were asked about their current and previous use of radiation-producing equipment. Information regarding their level of knowledge and application of radiation safety principles and practice standards was collected and analysed. Results: The use of radiation-producing equipment was evident in 94% of responding clinics (a combination of X-ray, CT and/or Nuclear Medicine Cameras). Of those with radiation-producing equipment, 94% indicated that they hold a radiation licence, 78% had never completed a certified radiation safety course and 19% of participants did not use a personal radiation monitor. In 14% of cases, radiation safety manuals or protocols were not available within clinics. Conclusions: The study has shown that knowledge and application of guidelines as provided by the Code of Practice for Radiation Protection in Veterinary Medicine (2009) is poorly adhered to. The importance of compliance with regulatory requirements is pivotal in minimising occupational exposure to ionising radiation in veterinary medicine, thus there is a need for increased education and training in the area. - Highlights: • Application of the Code of Practice for Veterinary Medicine is poorly adhered to. • Majority of veterinary clinics had not completed certified radiation safety course. • One-fifth of participants did not use personal radiation monitoring. • Increased education and training in area of radiation safety and protection required to generate compliance in clinics

IRSN opinion on nuclear safety within the frame of the National Debate on Energy Transition

International Nuclear Information System (INIS)

2013-06-01

This note aims at outlining that nuclear safety must be taken into consideration in the debate on energy transition, and more particularly major accidents which may have a radiological impact on populations and on the environment. It outlines that the nuclear sector is generally perceived as a source of innovation, a booster for scientific, industrial and economic excellence, and that reducing the share of nuclear energy could affect these abilities. It discusses the relationship between ageing and safety for the French nuclear reactor fleet. It addresses the issue of a possible accident and the necessary provision of compensation costs. It finally outlines that, even though the probability of severe accidents for reactors of new generation is low, improvement actions are still to be studied, notably regarding the organisational and human context

Application of demographic analysis to pedestrian safety : final report.

Science.gov (United States)

2017-04-01

In recent years, many departments of transportation in the US have invested additional resources to enhance : pedestrian safety. However, there is still a need to effectively and systematically address the pedestrian experience : in low-income areas....

Safety studies project on waste management. Final report. Chapters 2 and 3

International Nuclear Information System (INIS)

1985-01-01

The report presents, in summary form, a mode of procedure for accident analysis in nuclear waste management facilities. New instruments for safety analysis have been developed and tested. The report describes exemplary safety analyses with the new instrumentation. The safety analyses were carried out in surface systems, i.e. reprocessing and waste treatment systems, and in underground nuclear waste storage road and rail transport of radioactive materials have been investigated. (EF) [de

Japan reforms its nuclear safety

International Nuclear Information System (INIS)

Anon.

2013-01-01

The Fukushima Daiichi NPP accident deeply questioned the bases of nuclear safety and nuclear safety regulation in Japan. It also resulted in a considerable loss of public confidence in the safety of nuclear power across the world. Although the accident was caused by natural phenomena, institutional and human factors also largely contributed to its devastating consequences, as shown by the Japanese Diet's and Government's investigation reports. 'Both regulators and licensees were held responsible and decided to fully reconsider the existing approaches to nuclear safety. Consequently, the regulatory system underwent extensive reform based on the lessons learned from the accident,' Yoshihiro Nakagome, the President of Japan Nuclear Energy Safety Organisation, an ETSON member TSO, explains. (orig.)

Developing patient safety in dentistry.

Science.gov (United States)

Pemberton, M N

2014-10-01

Patient safety has always been important and is a source of public concern. Recent high profile scandals and subsequent reports, such as the Francis report into the failings at Mid Staffordshire, have raised those concerns even higher. Mortality and significant morbidity associated with the practice of medicine has led to many strategies to help improve patient safety, however, with its lack of associated mortality and lower associated morbidity, dentistry has been slower at systematically considering how patient safety can be improved. Recently, several organisations, researchers and clinicians have discussed the need for a patient safety culture in dentistry. Strategies are available to help improve patient safety in healthcare and deserve further consideration in dentistry.

High level radioactive wastes: Considerations on final disposal

International Nuclear Information System (INIS)

Ciallella, Norberto R.

2000-01-01

When at the beginnings of the decade of the 80 the National Commission on Atomic Energy (CNEA) in Argentina decided to study the destination of the high level radioactive wastes, was began many investigations, analysis and multidisciplinary evaluations that be origin to a study of characteristics never before carried out in Argentina. For the first time in the country was faced the study of an environmental eventual problem, several decades before that the problem was presented. The elimination of the high level radioactive wastes in the technological aspects was taken in advance, avoiding to transfer the problems to the future generations. The decision was based, not only in technical evaluations but also in ethical premises, since it was considered that the future generations may enjoy the benefits of the nuclear energy and not should be solve the problem. The CNEA in Argentina in 1980 decided to begin a feasibility study and preliminary engineering project for the construction of the final disposal of high level radioactive wastes

Key issues on safety design basis selection and safety assessment

International Nuclear Information System (INIS)

An, S.; Togo, Y.

1976-01-01

In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

77 FR 21311 - Locomotive Safety Standards

Science.gov (United States)

2012-04-09

... preparedness, alcohol and drug testing, locomotive engineer certification, and workplace safety. In 1980, FRA... Association (ATDA) Amtrak AAR Association of Railway Museums (ARM) Association of State Rail Safety Managers... Administration 49 CFR Parts 229 and 238 Locomotive Safety Standards; Final Rule #0;#0;Federal Register / Vol. 77...

Nuclear safety policy working group recommendations on nuclear propulsion safety for the space exploration initiative

Science.gov (United States)

Marshall, Albert C.; Lee, James H.; Mcculloch, William H.; Sawyer, J. Charles, Jr.; Bari, Robert A.; Cullingford, Hatice S.; Hardy, Alva C.; Niederauer, George F.; Remp, Kerry; Rice, John W.

1993-01-01

An interagency Nuclear Safety Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program. These recommendations, which are contained in this report, should facilitate the implementation of mission planning and conceptual design studies. The NSPWG has recommended a top-level policy to provide the guiding principles for the development and implementation of the SEI nuclear propulsion safety program. In addition, the NSPWG has reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. These recommendations should be useful for the development of the program's top-level requirements for safety functions (referred to as Safety Functional Requirements). The safety requirements and guidelines address the following topics: reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations.

Nuclear safety: an operational constraint or necessity

International Nuclear Information System (INIS)

Gauvenet, A.

1983-01-01

Different aspects of the nuclear safety in the operation of power stations are analysed. There is always a danger that safety is considered as a constraint at operator level, but it is essential that human factors and working conditions be taken into consideration [fr

Interim Storage of Spent Nuclear Fuel before Final Disposal in Germany - Regulator's view

International Nuclear Information System (INIS)

Arens, G.; Goetz, Ch.; Geupel, Sandra; Gmal, B.; Mester, W.

2014-01-01

For spent nuclear fuel management in Germany the concept of dry interim storage in dual purpose casks before direct disposal is applied. The Federal Office for Radiation Protection (BfS) is the competent authority for licensing of interim storage facilities. The competent authority for surveillance of operation is the responsible authority of the respective federal state (Land). Currently operation licenses for storage facilities have been granted for a storage time of 40 years and are based on safety demonstrations for all safety issues as safe enclosure, shielding, sub-criticality and decay heat removal under consideration of operation conditions. In addition, transportability of the casks for the whole storage period has to be provided. Due to current delay in site selection and exploration of a disposal site, an extension of the storage time beyond 40 years could be needed. This will cause appropriate actions by the licensee and the competent authorities as well. A brief description of the regulatory base of licensing and surveillance of interim storage is given from the regulators view. Furthermore the current planning for final disposal of spent nuclear fuel and high level waste and its interconnections between storage and disposal concepts are shortly explained. Finally the relevant aspects for licensing of extended storage time beyond 40 years will be discussed. Current activities on this issue, which have been initiated by the Federal Government, will be addressed. On the regulatory side a review and amendment of the safety guideline for interim storage of spent fuel has been performed and the procedure of periodic safety review is being implemented. A guideline for implementing an ageing management programme is available in a draft version. Regarding safety of long term storage a study focussing on the identification and evaluation of long term effects as well as gaps of knowledge has been finished in 2010. A continuation and update is currently underway

Safety analysis of fusion reactors pertaining to nuclear incidents and accidents. Final report

International Nuclear Information System (INIS)

Raeder, J.; Weller, A.; Wolf, R.; Jin, X.; Boccaccini, L.V.; Stieglitz, R.; Carloni, D.; Pistner, C.; Herb, J.

2013-11-01

The BfS gave the projekt partners IPP, KIT, Oeko-Institut e. V., and GRS the order to carry out a literature study on the topic of safety of fusion power plants regarding nuclear incidents and accidents. In the framework of this study the actual status of science and technology of the safety concept of fusion power plants should be determined and the applicability of the nuclear safety regulations hitherto developed for nuclear power plants checked. For future commercial fusion power plants today only conceptional designs exist. The most advanced conceptual study for a future fusion power plant is the European Power Plant Conceptual Study (PPCS) from the year 2005, which is based on the tokamak principle. In this study also fundamental aspects of the safety concept of nuclear fusion are treated. Hereby several different conceptual approaches are discussed, which differ among others also in the lay-out approaches relevant for the safety of a facility like for instance the choice of the breeding concept or the materials for the blanket/divertor structure and the coolants. The safety concept of nuclear fusion is oriented on safety concepts for facilities with radioactive inventory. It is based on the concept of tiered safety levels. In order to check whether for the nuclear fusion a safety concept comparable with the nuclear fission at all is necessary, in a first step it was considered, which consequences are possible at a postulated release o large parts of the radioactive inventory of a fusion power plant. Such a worst-case scenario was compared with a corresponding, postulated release of large parts of the radioactive inventory of a nuclear power plant. As scale hereby served the radiological criterion, at the transgression of which in the environment of the facility an evacuation would be necessary. In a next step the transferability of the safety concept of the tiered safety levels of nuclear technology to the fusion was checked. Beside events transferable from

Operational safety and radiation protection considerations in designing an HLW repository in Germany

International Nuclear Information System (INIS)

Filbert, W.; Kreienmeyer, M.; Poehler, M.; Niehues, N.

2008-01-01

In Germany the reference concept for disposal of heat generating radioactive waste considers emplacing canisters with vitrified waste in deep vertical boreholes drilled from the drifts of a repository mine in salt at a depth of 870 m. Spent fuel is to be disposed of in self-shielding POLLUX casks in horizontal drifts. An optimized disposal concept anticipates emplacing unshielded canisters with vitrified HLW and canisters containing the fuel rods of 3 PWR or 9 BWR fuel assemblies in boreholes with a diameter of 60 cm and a depth of up to 300 m.. In all cases the void space between POLLUX cask and drifts and canisters and borehole wall will be backfilled with crushed salt. (1) Operational Safety: Based on a detailed description of all underground disposal operation steps, the possible impacts on the disposal operations were analysed and the need for further studies determined. The disposal operation steps comprise e.g. rail bound transport from the shaft to the emplacement drift and emplacement process itself. As possible impacts the following occurrences were considered: ventilation failure, power supply failure, rock mechanics impact including cross-section convergence, irregular floor uplift and rock fall, brine and natural gas intrusion, derailing of transport carts and finally internal fire. (2) Radiation Protection: According to the German Atomic Energy Act (AtG), the design, construction and operation of a nuclear site like a final repository has to be licensed by the responsible authority. The Radiological Protection Ordinance and further guidelines i.e. concerning the emission and immission of released radioactive nuclides or the risk analysis of possible failure, build the basis for the licensing procedures. To ensure adequate protection against undue radiation exposure the repository is divided into different radiological protection areas. Generally, the handling of shielded waste packages above und under ground (including all the pathway of transport and

An interagency space nuclear propulsion safety policy for SEI - Issues and discussion

Science.gov (United States)

Marshall, A. C.; Sawyer, J. C., Jr.

1991-01-01

An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top level safety requirements and guidelines to address these issues. Safety topics include reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations. In this paper the emphasis is placed on the safety policy and the issues and considerations that are addressed by the NSPWG recommendations.

Safety climate in OHSAS 18001-certified organisations: antecedents and consequences of safety behaviour.

Science.gov (United States)

Fernández-Muñiz, Beatriz; Montes-Peón, José Manuel; Vázquez-Ordás, Camilo José

2012-03-01

The occupational health and safety standard OHSAS 18001 has gained considerable acceptance worldwide, and firms from diverse sectors and of varying sizes have implemented it. Despite this, very few studies have analysed safety management or the safety climate in OHSAS 18001-certified organisations. The current work aims to analyse the safety climate in these organisations, identify its dimensions, and propose and test a structural equation model that will help determine the antecedents and consequences of employees' safety behaviour. For this purpose, the authors carry out an empirical study using a sample of 131 OHSAS 18001-certified organisations located in Spain. The results show that management's commitment, and particularly communication, have an effect on safety behaviour and on safety performance, employee satisfaction, and firm competitiveness. These findings are particularly important for management since they provide evidence about the factors that should be encouraged to reduce risks and improve performance in this type of organisation. Copyright © 2011 Elsevier Ltd. All rights reserved.

Building the safety case for a hypothetical underground repository in crystalline rock. Final report. Vol. 2. Safety file

International Nuclear Information System (INIS)

Biurrun, E.; Engelmann, H.J.; Jobmann, M.; Lommerzheim, A.; Popp, W.; Frentz, R.R. v.; Wahl, A.

1996-10-01

The study was intended as a desk simulation of the process of preparing a licensing application for a deep repository for spent fuel and high level waste in crystalline rock. After clarifying of organizational aspects of table of contents specifying all aspects in a safety life for license application were considered. The volume II is subdivided in two parts. Part A describes the general information, waste description, site characteristics, disposal facility design, reporitory construction and operation, quality assurance, operational safety, repository closure, organization and financial aspects, and long-term safety assessment. Part B deals with the impact of retrievability. (DG)

Safety design requirements for safety systems and components of JSFR

International Nuclear Information System (INIS)

Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

2011-01-01

Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

CONSIDERATIONS ON SOILS ISOLATIVE PROPERTIES FOR SITING OF A NEW NEAR-SURFACE RADIOACTIVE WASTE REPOSITORY IN POLAND IN THE LIGHT OF THE LONG TERM SAFETY

Directory of Open Access Journals (Sweden)

Monika Skrzeczkowska

2012-07-01

Full Text Available The paper presents a brief description of the occurrence of favorable isolative conditions for new surface radioactive waste repository in Poland. Selected soils may be used as a natural bottom layer or engineering barrier in multi-barrier system of RW repository. Currently, there is no regulation establishing standards for the bottom isolation, and the only quantifiable parameter with regard to water permeability is given for the repository objects, which in their case has to be lower than 10-9 m/s. For the purposes of this paper, treating on providing suitable bottom isolation for the new repository, this parameter has been transferred onto the consideration for soils suitability with a statement that it shall not be lower than the one given for the infrastructure. Submitted information should be taken into consideration by updating the information for the siting process according to IAEA Safety Standards.

Safety analyses for reprocessing and waste processing

International Nuclear Information System (INIS)

1983-03-01

Presentation of an incident analysis of process steps of the RP, simplified considerations concerning safety, and safety analyses of the storage and solidification facilities of the RP. A release tree method is developed and tested. An incident analysis of process steps, the evaluation of the SRL-study and safety analyses of the storage and solidification facilities of the RP are performed in particular. (DG) [de

Space reactor safety, 1985--1995 lessons learned

International Nuclear Information System (INIS)

Marshall, A.C.

1995-01-01

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration

Space reactor safety, 1985--1995 lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marshall, A.C.

1995-12-31

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

Indicators of safety culture - selection and utilization of leading safety performance indicators

Energy Technology Data Exchange (ETDEWEB)

Reiman, Teemu; Pietikaeinen, Elina (VTT, Technical Research Centre of Finland (Finland))

2010-03-15

Safety indicators play a role in providing information on organizational performance, motivating people to work on safety and increasing organizational potential for safety. The aim of this report is to provide an overview on leading safety indicators in the domain of nuclear safety. The report explains the distinction between lead and lag indicators and proposes a framework of three types of safety performance indicators - feedback, monitor and drive indicators. Finally the report provides guidance for nuclear energy organizations for selecting and interpreting safety indicators. It proposes the use of safety culture as a leading safety performance indicator and offers an example list of potential indicators in all three categories. The report concludes that monitor and drive indicators are so called lead indicators. Drive indicators are chosen priority areas of organizational safety activity. They are based on the underlying safety model and potential safety activities and safety policy derived from it. Drive indicators influence control measures that manage the socio technical system; change, maintain, reinforce, or reduce something. Monitor indicators provide a view on the dynamics of the system in question; the activities taking place, abilities, skills and motivation of the personnel, routines and practices - the organizational potential for safety. They also monitor the efficacy of the control measures that are used to manage the socio technical system. Typically the safety performance indicators that are used are lagging (feedback) indicators that measure the outcomes of the socio technical system. Besides feedback indicators, organizations should also acknowledge the important role of monitor and drive indicators in managing safety. The selection and use of safety performance indicators is always based on an understanding (a model) of the socio technical system and safety. The safety model defines what risks are perceived. It is important that the safety

Indicators of safety culture - selection and utilization of leading safety performance indicators

International Nuclear Information System (INIS)

Reiman, Teemu; Pietikaeinen, Elina

2010-03-01

Safety indicators play a role in providing information on organizational performance, motivating people to work on safety and increasing organizational potential for safety. The aim of this report is to provide an overview on leading safety indicators in the domain of nuclear safety. The report explains the distinction between lead and lag indicators and proposes a framework of three types of safety performance indicators - feedback, monitor and drive indicators. Finally the report provides guidance for nuclear energy organizations for selecting and interpreting safety indicators. It proposes the use of safety culture as a leading safety performance indicator and offers an example list of potential indicators in all three categories. The report concludes that monitor and drive indicators are so called lead indicators. Drive indicators are chosen priority areas of organizational safety activity. They are based on the underlying safety model and potential safety activities and safety policy derived from it. Drive indicators influence control measures that manage the socio technical system; change, maintain, reinforce, or reduce something. Monitor indicators provide a view on the dynamics of the system in question; the activities taking place, abilities, skills and motivation of the personnel, routines and practices - the organizational potential for safety. They also monitor the efficacy of the control measures that are used to manage the socio technical system. Typically the safety performance indicators that are used are lagging (feedback) indicators that measure the outcomes of the socio technical system. Besides feedback indicators, organizations should also acknowledge the important role of monitor and drive indicators in managing safety. The selection and use of safety performance indicators is always based on an understanding (a model) of the socio technical system and safety. The safety model defines what risks are perceived. It is important that the safety

Energy. Health, environment, and safety hazards. Final report from the Energy Commission. Energi. Haelso- miljoe- och saekerhetsrisker. Slutbetaenkande av energikommissionen

Energy Technology Data Exchange (ETDEWEB)

1978-01-01

The Swedish Energy Commission in its main report (''Energy'', SOU 1978:17) presented its considerations and put forward its proposals for a Swedish Energy policy for the next decade. This report contains complementary information on health hazards, risks of major accidents and sabotage, and problems of waste management. The presentation takes the form of a comparison of such risks in relation to different sources of energy. The Commission is not unanimous in its estimates of the relative hazards of different energysystems. The Commission recommends the initiation of a large number of studies concerning the possible ways the increase the safety and reduce the adverse effects of energy production.

Considerations about radiological safety in industrial gammagraphy

International Nuclear Information System (INIS)

Sajaroff, P.M.

1988-01-01

Gammagraphy is the most dangerous practice out of the commonly employed with radioactive materials or radiation sources in industry, medicine, research and training. The amount of accidents which took place in different countries and their radiological consequences oblige to take extreme precautions in order to reduce their happening to the maximum. The principal cause of the accidents is the human error, the most common of which are described. Safe operation depends on the quality of the equipment and the human factors, thus it is mandatory to accomplish a series of requirements, the most important of which are presented. The solution of the problem mainly requires the application of an adequate programme of quality warranty to the design, construction, testing and operation. Furthermore, the radiation sources must be identified by unnoticed persons. Finally a brief description of some accidents which took place in different countries are presented. (M.E.L.) [es

Local Anesthetics: Review of Pharmacological Considerations

Science.gov (United States)

Becker, Daniel E; Reed, Kenneth L

2012-01-01

Local anesthetics have an impressive history of efficacy and safety in medical and dental practice. Their use is so routine, and adverse effects are so infrequent, that providers may understandably overlook many of their pharmacotherapeutic principles. The purpose of this continuing education article is to provide a review and update of essential pharmacology for the various local anesthetic formulations in current use. Technical considerations will be addressed in a subsequent article. PMID:22822998

A feasibility study for Arizona's roadway safety management process using the Highway Safety Manual and SafetyAnalyst : final report.

Science.gov (United States)

2016-07-01

To enable implementation of the American Association of State Highway Transportation (AASHTO) Highway Safety Manual using : SaftetyAnalyst (an AASHTOWare software product), the Arizona Department of Transportation (ADOT) studied the data assessment :...

Introduction to safety theory

International Nuclear Information System (INIS)

Meyna, A.

1982-01-01

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

The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

International Nuclear Information System (INIS)

Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

2006-06-01

Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

Engineering nucleases for gene targeting: safety and regulatory considerations.

Science.gov (United States)

Pauwels, Katia; Podevin, Nancy; Breyer, Didier; Carroll, Dana; Herman, Philippe

2014-01-25

Nuclease-based gene targeting (NBGT) represents a significant breakthrough in targeted genome editing since it is applicable from single-celled protozoa to human, including several species of economic importance. Along with the fast progress in NBGT and the increasing availability of customized nucleases, more data are available about off-target effects associated with the use of this approach. We discuss how NBGT may offer a new perspective for genetic modification, we address some aspects crucial for a safety improvement of the corresponding techniques and we also briefly relate the use of NBGT applications and products to the regulatory oversight. Copyright © 2013 Elsevier B.V. All rights reserved.

Safety culture in nuclear power plants

International Nuclear Information System (INIS)

Weihe, G. von; Pamme, H.

2003-01-01

Experience shows that German nuclear power plants have always been operated reliably and safely. Over the years, the safety level in these plants has been raised considerably so that they can stand any comparison with other countries. This is confirmed by the two reports published by the Federal Ministry for the Environment on the nuclear safety convention. Behind this, there must obviously stand countless appropriate 'good practices' and a safety management system in nuclear power plants. (orig.) [de

Engineering considerations for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Scully, L.W.

1978-01-01

The WIPP, located at Los Medanos in New Mexico, is to be used for DOE transuranic and high-level defense wastes. On the surface, there are contact-handled and remote-handled waste facilities. Package size, delivery rates, shipping, shielding and thermal considerations, underground transport and emplacement, retrievability, ventilation, and hoist conveyence safety are discussed

CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FORTHE DECOMMISSIONING OF THE FUEL CHANNELS IN THECANDU NUCLEAR REACTOR

Directory of Open Access Journals (Sweden)

Gabi ROSCA FARTAT

2013-05-01

Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of theseinstallations has become one of the 21stcentury’s great challenges. Each project may be managed differently,depending on the country, development policies, financial considerations, and the availability of qualifiedengineers or specialized companies to handle such projects. The principle objective of decommissioning is toplace a facility into such a condition that there is no unacceptable risk from the decommissioned facility topublic health and safety of the environment. In order to ensure that at the end of its life the risk from a facility iswithin acceptable bounds, action is normally required. The overall decommissioning strategy is todeliver a timely, cost-effective program while maintaining high standards of safety, security and environmentalprotection. If facilities were not decommissioned, they could degrade and potentially present an environmentalradiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is notconsidered to be an acceptable alternative to decommissioning. The final aim of decommissioning is torecover the geographic site to its original condition.

Planning the Unplanned Experiment: Towards Assessing the Efficacy of Standards for Safety-Critical Software

Science.gov (United States)

Graydon, Patrick J.; Holloway, C. M.

2015-01-01

Safe use of software in safety-critical applications requires well-founded means of determining whether software is fit for such use. While software in industries such as aviation has a good safety record, little is known about whether standards for software in safety-critical applications 'work' (or even what that means). It is often (implicitly) argued that software is fit for safety-critical use because it conforms to an appropriate standard. Without knowing whether a standard works, such reliance is an experiment; without carefully collecting assessment data, that experiment is unplanned. To help plan the experiment, we organized a workshop to develop practical ideas for assessing software safety standards. In this paper, we relate and elaborate on the workshop discussion, which revealed subtle but important study design considerations and practical barriers to collecting appropriate historical data and recruiting appropriate experimental subjects. We discuss assessing standards as written and as applied, several candidate definitions for what it means for a standard to 'work,' and key assessment strategies and study techniques and the pros and cons of each. Finally, we conclude with thoughts about the kinds of research that will be required and how academia, industry, and regulators might collaborate to overcome the noted barriers.

Public Health Service Safety Program

Energy Technology Data Exchange (ETDEWEB)

McBride, J R [Southwestern Radiological Health Laboratory, Las Vegas, NV (United States)

1969-07-01

Off-Site Radiological Safety Programs conducted on past Plowshare experimental projects by the Southwestern Radiological Health Laboratory for the AEC will be presented. Emphasis will be placed on the evaluation of the potential radiation hazard to off-site residents, the development of an appropriate safety plan, pre- and post-shot surveillance activities, and the necessity for a comprehensive and continuing community relations program. In consideration of the possible wide use of nuclear explosives in industrial applications, a new approach to off-site radiological safety will be discussed. (author)

Public Health Service Safety Program

International Nuclear Information System (INIS)

McBride, J.R.

1969-01-01

Off-Site Radiological Safety Programs conducted on past Plowshare experimental projects by the Southwestern Radiological Health Laboratory for the AEC will be presented. Emphasis will be placed on the evaluation of the potential radiation hazard to off-site residents, the development of an appropriate safety plan, pre- and post-shot surveillance activities, and the necessity for a comprehensive and continuing community relations program. In consideration of the possible wide use of nuclear explosives in industrial applications, a new approach to off-site radiological safety will be discussed. (author)

Integrated Safety in Design

DEFF Research Database (Denmark)

Schultz, Casper Siebken; Jørgensen, Kirsten

2014-01-01

An on-going research project investigates the inclusion of health and safety considerations in the design phase as a means to achieve a higher level of health and safety in the construction industry. Moreover, the approach is coupled to the overall quality efforts. Two architectural firms and two...... consulting engineering firms are project participants. The hypothesis is that health and safety problems in execution can be prevented through better planning in the early stages of the construction processes and that accidents are prevented by providing safety. In the first stage of the research project...... a theoretical framework is developed from a combination of existing literature on health and safety and a mapping of existing practices based on interviews in all four companies. The interviews revealed that the basic knowledge on OHS among architects and engineers is limited. Also currently designers typically...

Conduct and results of the Interagency Nuclear Safety Review Panel's evaluation of the Ulysses space mission

International Nuclear Information System (INIS)

Sholtis, J.A. Jr.; Gray, L.B.; Huff, D.A.; Klug, N.P.; Winchester, R.O.

1991-01-01

The recent 6 October 1990 launch and deployment of the nuclear-powered Ulysses spacecraft from the Space Shuttle Discovery culminated an extensive safety review and evaluation effort by the Interagency Nuclear Safety Review Panel (INSRP). After more than a year of detailed independent review, study, and analysis, the INSRP prepared a Safety Evaluation Report (SER) on the Ulysses mission, in accordance with Presidential Directive-National Security Council memorandum 25. The SER, which included a review of the Ulysses Final Safety Analysis Report (FSAR) and an independent characterization of the mission risks, was used by the National Aeronautics and Space Administration (NASA) in its decision to request launch approval as well as by the Executive Office of the President in arriving at a launch decision based on risk-benefit considerations. This paper provides an overview of the Ulysses mission and the conduct as well as the results of the INSRP evaluation. While the mission risk determined by the INSRP in the SER was higher than that characterized by the Ulysses project in the FSAR, both reports indicated that the radiological risks were relatively small. In the final analysis, the SER proved to be supportive of a positive launch decision. The INSRP evaluation process has demonstrated its effectiveness numerous times since the 1960s. In every case, it has provided the essential ingredients and perspective to permit an informed launch decision at the highest level of our Government

The Pit and the Safety Pendulum

Energy Technology Data Exchange (ETDEWEB)

Nitschke, Robert Leon; Ramos, Amadeo Gabriel

2000-11-01

The hypothesis of this paper is that the safety analysis pendulum has swung considerably in the direction of increasingly complex and lengthy safety evaluations and intense reviews during the past 30 years. The test of this hypothesis will be a review of the safety analysis conducted for various activities associated with the retrieval of transuranic radioactive waste from burial pits at a National Laboratory site over a span of 30 years. The examination will focus on the safety aspects and the safety analysis that was conducted for the projects. At the conclusion of this examination, the paper will identify five reasons why the changes have taken place.

A Review of General Aviation Safety (1984-2017).

Science.gov (United States)

Boyd, Douglas D

2017-07-01

General aviation includes all civilian aviation apart from operations involving paid passenger transport. Unfortunately, this category of aviation holds a lackluster safety record, accounting for 94% of civil aviation fatalities. In 2014, of 1143 general aviation accidents, 20% were fatal compared with 0 of 29 airline mishaps in the United States. Herein, research findings over the past 30 yr will be reviewed. Accident risk factors (e.g., adverse weather, geographical region, post-impact fire, gender differences) will be discussed. The review will also summarize the development and implementation of stringent crashworthiness designs with multi-axis dynamic testing and head-injury protection and its impact on mitigating occupant injury severity. The benefits and drawbacks of new technology and human factor considerations associated with increased general aviation automation will be debated. Data on the safety of the aging general aviation population and increased drug usage will also be described. Finally, areas in which general aviation occupant survival could be improved and injury severity mitigated will be discussed with the view of equipping aircraft with 1) crash-resistant fuel tanks to reduce post-impact conflagration; 2) after-market ballistic parachutes for older aircraft; and 3) current generation electronic locator beacons to hasten site access by first responders.Boyd DD. A review of general aviation safety (1984-2017). Aerosp Med Hum Perform. 2017; 88(7):657-664.

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of the elastic-plastic material properties (Phase 2). Final report; Erhoehung der Komponentensicherheit durch verbesserte Verfahren zur Eigenspannungsanalyse. Teilvorhaben. Beruecksichtigung der elastisch-plastischen Materialeigenschaften (Phase 2). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Mirbach, David von

2015-12-16

Residual stresses in mechanical components can result in both detrimental but also beneficial effects on the strength and lifetime of the components. The most detailed knowledge of the residual stress state is of advantage or a pre-requisite for the assessment of the component performance. Two commonly used methods for determination of residual stresses are the hole drilling method and the ring core method which can be regarded to the mechanical methods. In the context of reactor safety research of the German Federal Ministry of Economic and Energy (BMWi) two fundamental and interacting weak points of the hole drilling method as well as of the ring core method, respectively, in order to determine residual stresses are going to be investigated. As a consequence reliability of the methods will be improved in this joint research project. On the one hand there are effects of geometrical boundary conditions of the components and on the other hand there is the influence of plasticity due to notch effects both affecting the released strain field after removing material and after all the calculated residual stresses. The first issue mentioned above is under the responsibility of the Institute of Materials Engineering (Kassel University) and the last one is investigated by materials testing institute university Stuttgart. As a consequence of a successful project the knowledge base will be considerably improved resulting in benefits for various engineering fields. Especially the quantitative consideration of real residual stress states for optimized component designs will be possible and after all the consequences of residual stresses on safety of components which are used in nuclear facilities can be evaluated. In this second experimental research chapter (phase 2) the findings of the first numerical and theoretical research chapter (phase 1) where proofed. The developed differential calculation method with the method of adaptive calibration functions were compared with the

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of the elastic-plastic material properties (phase 1). Final report; Erhoehung der Komponentensicherheit durch verbesserte Verfahren zur Eigenspannungsanalyse. Teilvorhaben. Beruecksichtigung der elastisch-plastischen Materialeigenschaften (Phase 1). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Mirbach, David von

2014-07-28

Residual stresses in mechanical components can result in both detrimental but also beneficial effects on the strength and lifetime of the components. The most detailed knowledge of the residual stress state is of advantage or a pre-requisite for the assessment of the component performance. The mechanical methods for residual stress measurement are divided into the groups of non-destructive and destructive methods. Two commonly used mechanical methods for determination of residual stresses are the hole drilling method and the ring core method which can be regarded as semi-destructive methods. In the context of reactor safety research of the German Federal Ministry of Economic and Technology (BMWi) two fundamental and interacting weak points of the hole drilling method as well as of the ring core method, respectively, in order to determine residual stresses are going to be investigated. As a consequence reliability of the methods will be improved in this joint research project. On the one hand there are effects of geometrical boundary conditions of the components and on the other hand there is the influence of plasticity due to notch effects both affecting the released strain field after removing material and after all the calculated residual stresses. The first issue mentioned above is under the responsibility of the Institute of Materials Engineering (Kassel University) and the last one is investigated by Universitaet of Stuttgart-Otto-Graf-Institut - materials testing institute. As a consequence of a successful project the knowledge base will be considerably improved resulting in benefits for various engineering fields. Especially the quantitative consideration of real residual stress states for optimized component designs will be possible and after all the consequences of residual stresses on safety of components which are used in nuclear facilities can be evaluated. The state of art was reground in the first research chapter and the analysed strain gauges where

Technical considerations in the design of near surface disposal facilities for radioactive waste

International Nuclear Information System (INIS)

2001-11-01

Good design is an important step towards ensuring operational as well as long term safety of low and intermediate level waste (LILW) disposal. The IAEA has produced this report with the objective of outlining the most important technical considerations in the design of near surface disposal facilities and to provide some examples of the design process in different countries. This guidance has been developed in light of experience gained from the design of existing near surface disposal facilities in a range of Member States. In particular the report provide information on design objective, design requirements, and design phases. The report focuses on: near surface disposal facilities accepting solidified LILW; disposal facilities on or just below the ground surface, where the final protective covering is of the order of a few metres thick; and disposal facilities several tens of metres below the ground surface (including rock cavern type facilities)

Radiological and environmental consequences. Final report of the Nordic Nuclear Safety Research project BOK-2

International Nuclear Information System (INIS)

Palsson, S.E.

2002-11-01

Final report of the Nordic Nuclear Safety Research project BOK-2, Radiological and Environmental Consequences. The project was carried out 1998-2001 with participants from all the Nordic countries. Representatives from the Baltic States were also invited to some of the meetings and seminars. The project consisted of work on terrestrial and marine radioecology and had a broad scope in order to enable participation of research groups with various fields of interest. This report focuses on the project itself and gives a general summary of the studies undertaken. A separate technical report summarises the work done by each research group and gives references to papers published in scientific journals. The topics in BOK-2 included improving assessment of old and recent fallout, use of radionuclides as tracers in Nordic marine areas, improving assessment of internal doses and use of mass spectrometry in radioecology. (au)

Considerations about the licensing process of special nuclear industrial facilities

Energy Technology Data Exchange (ETDEWEB)

Talarico, M.A., E-mail: talaricomarco@hotmail.com [Marinha do Brasil, Rio de Janeiro, RJ (Brazil). Coordenacao do Porgrama de Submarino com Propulsao Nuclear; Melo, P.F. Frutuoso e [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2015-07-01

This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

Considerations about the licensing process of special nuclear industrial facilities

International Nuclear Information System (INIS)

Talarico, M.A.; Melo, P.F. Frutuoso e

2015-01-01

This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

Development of Taiwan's strategies for regulating nanotechnology-based pharmaceuticals harmonized with international considerations.

Science.gov (United States)

Guo, Jiun-Wen; Lee, Yu-Hsuan; Huang, Hsiau-Wen; Tzou, Mei-Chyun; Wang, Ying-Jan; Tsai, Jui-Chen

2014-01-01

Nanotechnology offers potential in pharmaceuticals and biomedical developments for improving drug delivery systems, medical imaging, diagnosis, cancer therapy, and regenerative medicine. Although there is no international regulation or legislation specifically for nanomedicine, it is agreed worldwide that considerably more attention should be paid to the quality, safety, and efficacy of nanotechnology-based drugs. The US Food and Drug Administration and the European Medicines Agency have provided several draft regulatory guidance and reflection papers to assist the development of nanomedicines. To cope with the impact of nanotechnology and to foster its pharmaceutical applications and development in Taiwan, this article reviews the trends of regulating nanotechnology-based pharmaceuticals in the international community and proposes strategies for Taiwan's regulation harmonized with international considerations. The draft regulatory measures include a chemistry, manufacturing, and controls (CMC) review checklist and guidance for CMC review of liposomal products. These have been submitted for discussion among an expert committee, with membership comprised of multidisciplinary academia, research institutions, the pharmaceutical industry, and regulators, and are currently approaching final consensus. Once a consensus is reached, these mechanisms will be recommended to the Taiwan Food and Drug Administration for jurisdiction and may be initiated as the starting point for regulating nanotechnology-based pharmaceuticals in Taiwan.

Development and Execution of the RUNSAFE Runway Safety Bayesian Belief Network Model

Science.gov (United States)

Green, Lawrence L.

2015-01-01

One focus area of the National Aeronautics and Space Administration (NASA) is to improve aviation safety. Runway safety is one such thrust of investigation and research. The two primary components of this runway safety research are in runway incursion (RI) and runway excursion (RE) events. These are adverse ground-based aviation incidents that endanger crew, passengers, aircraft and perhaps other nearby people or property. A runway incursion is the incorrect presence of an aircraft, vehicle or person on the protected area of a surface designated for the landing and take-off of aircraft; one class of RI events simultaneously involves two aircraft, such as one aircraft incorrectly landing on a runway while another aircraft is taking off from the same runway. A runway excursion is an incident involving only a single aircraft defined as a veer-off or overrun off the runway surface. Within the scope of this effort at NASA Langley Research Center (LaRC), generic RI, RE and combined (RI plus RE, or RUNSAFE) event models have each been developed and implemented as a Bayesian Belief Network (BBN). Descriptions of runway safety issues from the literature searches have been used to develop the BBN models. Numerous considerations surrounding the process of developing the event models have been documented in this report. The event models were then thoroughly reviewed by a Subject Matter Expert (SME) panel through multiple knowledge elicitation sessions. Numerous improvements to the model structure (definitions, node names, node states and the connecting link topology) were made by the SME panel. Sample executions of the final RUNSAFE model have been presented herein for baseline and worst-case scenarios. Finally, a parameter sensitivity analysis for a given scenario was performed to show the risk drivers. The NASA and LaRC research in runway safety event modeling through the use of BBN technology is important for several reasons. These include: 1) providing a means to clearly

The effect of Health, Safety and Environment Management System (HSE-MS on the improvement of safety performance indices in Urea and Ammonia Kermanshah Petrochemical Company

Directory of Open Access Journals (Sweden)

M. S. Poursoleiman

2015-09-01

.Conclusion: The implementation of Health, Safety and the Environment Management System caused a reduction in accidents and its consequences and most of the safety performance indices in the entire process cycle of Kermanshah Petrochemical Company. Overall, safety condition has been improved considerably.

Selected environmental considerations and their measuring parameters for nuclear power plant siting

International Nuclear Information System (INIS)

Norris, J.A.

1975-01-01

The site selection process for nuclear power stations encompasses a broad range of considerations. A categorization of these considerations consistent with the needs of the U. S. Atomic Energy Commission, as the regulatory agency, and of the utility company involves these major areas of concern. They are issues related to safety, environmental impact, and engineering/economics. The more important environmental considerations and their measuring parameters presented in this paper include biota, ecological systems and water quality, land use, aesthetics, water availability, and meteorology. (U.S.)

Evaluation of safety-parameter display concepts. Final report

International Nuclear Information System (INIS)

Woods, D.D.; Wise, J.A.; Hanes, L.F.

1982-02-01

New control room equipment designed to improve operator performance must be evaluated before adoption and installation. Two experimental concepts for a Safety Parameters Display System (SPDS) were evaluated to assess benefits and potential problems associated with the SPDS concept and its integration into control room operations. Participants were licensed utility operators undergoing retraining on a nuclear power plant simulator. Both quantitative and qualitative data were collected and analyzed on crew response to seven simulated accident conditions

Political economy and social psychology of nuclear safety

International Nuclear Information System (INIS)

Choe, Gwang Sik

2009-03-01

The contents of this book are consideration on independence of nuclear safety regulations, analysis of trend in internal and external on effectualness of nuclear safety regulations, political psychology of a hard whistle, how to deal with trust and distrust on regulation institute, international trend and domestic trend of nuclear safe culture, policy for building of trust of people on nuclear safety and regulations, measurement and conception of nuclear safety and for who imposes legal controls?