Tangential stretching rate (TSR) analysis of non premixed reactive flows

KAUST Repository

Valorani, Mauro

2016-10-16

We discuss how the Tangential stretching rate (TSR) analysis, originally developed and tested for spatially homogeneous systems (batch reactors), is extended to spatially non homogeneous systems. To illustrate the effectiveness of the TSR diagnostics, we study the ignition transient in a non premixed, reaction–diffusion model in the mixture fraction space, whose dependent variables are temperature and mixture composition. The reactive mixture considered is syngas/air. A detailed H2/CO mechanism with 12 species and 33 chemical reactions is employed. We will discuss two cases, one involving only kinetics as a model of front propagation purely driven by spontaneous ignition, the other as a model of deflagration wave involving kinetics/diffusion coupling. We explore different aspects of the system dynamics such as the relative role of diffusion and kinetics, the evolution of kinetic eigenvalues, and of the tangential stretching rates computed by accounting for the combined action of diffusion and kinetics as well for kinetics only. We propose criteria based on the TSR concept which allow to identify the most ignitable conditions and to discriminate between spontaneous ignition and deflagration front.

Tumor-stroma ratio(TSR) as a potential novel predictor of prognosis in digestive system cancers: A meta-analysis.

Science.gov (United States)

Zhang, Runjin; Song, Wei; Wang, Kai; Zou, Shubing

2017-09-01

The tumor-stroma ratio (TSR) has been reported as a prognosis predictor in multiple cancers. The aim of this meta-analysis was to investigate the potential value of TSR as a prognostic predictor of cancer in the digestive system. We searched PubMed, Embase, Elsevier and Web of Science. All studies exploring the association of TSR with overall survival (OS) or disease-free survival (DFS), and lymph node metastasis (LNM) were identified. In total, eight studies were eligible for analysis, and they included 1959 patients. Meta-analysis showed that the low TSR in the tumor could predict poor overall survival (OS) in multiple cancers (pooled Hazard Ratio [HR]: 2.15, 95%CI: 1.80-2.57, P<0.00001, fixed effects). For disease-free survival (DFS), low TSR was also a significant predictor (pooled Hazard Ratio [HR]: 2.31, 95%CI: 1.88-2.83, P<0.00001, fixed effects). In addition, low TSR was correlated with tumor stage. The tumor-stroma ratio (TSR) may potentially serve as a poor prognostic predictor for the metastasis and prognosis of cancer. Copyright © 2017. Published by Elsevier B.V.

Technical Safety Requirements for the Waste Storage Facilities

International Nuclear Information System (INIS)

Laycak, D.T.

2010-01-01

This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2009). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas, consisting

Technical Safety Requirements for the Waste Storage Facilities

Energy Technology Data Exchange (ETDEWEB)

Laycak, D T

2008-06-16

This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the 'Documented Safety Analysis for the Waste Storage Facilities' (DSA) (LLNL 2008). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas

HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

International Nuclear Information System (INIS)

EVANS, C.B.

2004-01-01

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

HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

Energy Technology Data Exchange (ETDEWEB)

EVANS, C B

2004-12-21

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

Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

Energy Technology Data Exchange (ETDEWEB)

1994-06-20

This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D&D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities.

Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

International Nuclear Information System (INIS)

1994-01-01

This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D ampersand D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities

Aspects of safety assessments for package with additional equipment components

International Nuclear Information System (INIS)

Reiche, I.; Boerst, F.M.; Krietsch, T.

2004-01-01

Many paragraphs in TS-R-1 contain the terms ''package'' or ''packaging''. These terms are defined in TS-R-1 paras 230 and 231 and explained in TS-G-1.1 paras 230.1 - 230.6. The importance of a consistent understanding of these definitions has been shown by recent discussions during the assessment of applications for package design approval. There was disagreement, if equipment components attached to the container body during transport, e.g. a transport frame, should be considered part of the package and taken into account in the safety assessment for the package. Discussions were also caused by the way inner design components are treated in the safety assessment of the package. This paper summarises the regulatory requirements to such additional equipment components and presents the way of their inclusion into the package design approval process in Germany

Coordination and collaboration between National Regulators for the Safety and Security of International Shipments of Radioactive Materials

International Nuclear Information System (INIS)

Whittingham, Stephen

2016-01-01

The safety record of transporting radioactive material is remarkable; globally with tens of millions of packages transported; each year with approximately 2-3% is related to the nuclear industry. Much of this success is due to the ongoing commitment made by the IAEA and its Member States to maintain over the past 50 years the prescriptive regulations for the transport of radioactive material (TS-R-1) and its associated guidance documents. TS-R-1 is by far the most popular IAEA document in terms of downloads and sales with some Member States adopting them directly into their domestic legal framework whilst others adopt them due to all of the TS-R-1 requirements being incorporated into the UN Model Regulations

Site-specific and synergistic stimulation of methylation on the bacterial chemotaxis receptor Tsr by serine and CheW

Directory of Open Access Journals (Sweden)

Weis Robert M

2005-03-01

Full Text Available Abstract Background Specific glutamates in the methyl-accepting chemotaxis proteins (MCPs of Escherichia coli are modified during sensory adaptation. Attractants that bind to MCPs are known to increase the rate of receptor modification, as with serine and the serine receptor (Tsr, which contributes to an increase in the steady-state (adapted methylation level. However, MCPs form ternary complexes with two cytoplasmic signaling proteins, the kinase (CheA and an adaptor protein (CheW, but their influences on receptor methylation are unknown. Here, the influence of CheW on the rate of Tsr methylation has been studied to identify contributions to the process of adaptation. Results Methyl group incorporation was measured in a series of membrane samples in which the Tsr molecules were engineered to have one available methyl-accepting glutamate residue (297, 304, 311 or 493. The relative rates at these sites (0.14, 0.05, 0.05 and 1, respectively differed from those found previously for the aspartate receptor (Tar, which was in part due to sequence differences between Tar and Tsr near site four. The addition of CheW generated unexpectedly large and site-specific rate increases, equal to or larger than the increases produced by serine. The increases produced by serine and CheW (added separately were the largest at site one, ~3 and 6-fold, respectively, and the least at site four, no change and ~2-fold, respectively. The rate increases were even larger when serine and CheW were added together, larger than the sums of the increases produced by serine and CheW added separately (except site four. This resulted in substantially larger serine-stimulated increases when CheW was present. Also, CheW enhanced methylation rates when either two or all four sites were available. Conclusion The increase in the rate of receptor methylation upon CheW binding contributes significantly to the ligand specificity and kinetics of sensory adaptation. The synergistic effect of

System Design and the Safety Basis

International Nuclear Information System (INIS)

Ellingson, Darrel

2008-01-01

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

The role of the Staphylococcal VraTSR regulatory system on vancomycin resistance and vanA operon expression in vancomycin-resistant Staphylococcus aureus.

Science.gov (United States)

Qureshi, Nadia K; Yin, Shaohui; Boyle-Vavra, Susan

2014-01-01

Vancomycin is often the preferred treatment for invasive methicillin-resistant Staphylococcus aureus (MRSA) infection. With the increase in incidence of MRSA infections, the use of vancomycin has increased and, as feared, isolates of vancomycin-resistant Staphylococcus aureus (VRSA) have emerged. VRSA isolates have acquired the entercoccal vanA operon contained on transposon (Tn) 1546 residing on a conjugal plasmid. VraTSR is a vancomycin and β-lactam-inducible three-component regulatory system encoded on the S. aureus chromosome that modulates the cell-wall stress response to cell-wall acting antibiotics. Mutation in vraTSR has shown to increase susceptibility to β-lactams and vancomycin in clinical VISA strains and in recombinant strain COLVA-200 which expresses a plasmid borne vanA operon. To date, the role of VraTSR in vanA operon expression in VRSA has not been demonstrated. In this study, the vraTSR operon was deleted from the first clinical VRSA strain (VRS1) by transduction with phage harvested from a USA300 vraTSR operon deletion strain. The absence of the vraTSR operon and presence of the vanA operon were confirmed in the transductant (VRS1Δvra) by PCR. Broth MIC determinations, demonstrated that the vancomycin MIC of VRS1Δvra (64 µg/ml) decreased by 16-fold compared with VRS1 (1024 µg/ml). The effect of the vraTSR operon deletion on expression of the van gene cluster (vanA, vanX and vanR) was examined by quantitative RT-PCR using relative quantification. A 2-5-fold decreased expression of the vanA operon genes occured in strain VRS1Δvra at stationary growth phase compared with the parent strain, VRS1. Both vancomycin resistance and vancomycin-induced expression of vanA and vanR were restored by complementation with a plasmid harboring the vraTSR operon. These findings demonstrate that expression in S. aureus of the horizontally acquired enterococcal vanA gene cluster is enhanced by the staphylococcal three-component cell wall stress regulatory

TSR: A Storage Ring for HIE-ISOLDE

CERN Document Server

Butler, P A; Blaum, K; Grieser, M; Davinson, T; Woods, P J; Flanagan, K; Freeman, S J; Lazarus, I H; Litvinov, Yu A; Raabe, R; Siesling, E; Wenander, F

2016-01-01

It is planned to install the heavy-ion, low-energy ring TSR, currently at the Max-Planck-Institute for Nuclear Physics in Heidelberg, at the HIE-ISOLDE facility in CERN, Geneva. Such a facility will provide a capability for experiments with stored, cooled secondary beams that is rich and varied, spanning from studies of nuclear ground-state properties and reaction studies of astrophysical relevance, to investigations with highly-charged ions and pure isomeric beams. In addition to experiments performed using beams recirculating within the ring, the cooled beams can be extracted and exploited by external spectrometers for high-precision measurements. The capabilities of the ring facility as well as some physics cases will be presented, together with a brief report on the status of the project.

An induction accelerator for the Heidelberg Test Storage Ring TSR

International Nuclear Information System (INIS)

Ellert, C.; Habs, D.; Music, M.; Schwalm, D.; Wolf, A.; Jaeschke, E.; Kambara, T.; Sigray, P.

1992-01-01

An induction accelerator has been installed in the heavy ion test storage ring TSR in Heidelberg. It allows for constant acceleration or deceleration of stored coasting ion beams without affecting their velocity profile and is well suited for ion beam manipulation in cooling experiments and for measurements of velocity dependent cooling forces. The design and operation of the device and first applications to laser cooling and to measurements of laser and electron cooling forces are described. (orig.)

Safety of Research Reactors. Safety Requirements

International Nuclear Information System (INIS)

2010-01-01

The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

Australian Experience in Implementing Transport Safety Regulations and Transport Security Recommendations

International Nuclear Information System (INIS)

Sarkar, S.

2016-01-01

Australian transport safety and security regulatory framework is governed by Commonwealth, State and Territory legislations. There are eleven competent authorities in Australia that includes three Commonwealth authorities, six states and two territory authorities. IAEA Regulations for Safe Transport of Radioactive Material (TS-R-1, 2005 edition) is applied through Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) Code of Practice for Transport of Radioactive Material 2008 by road, rail and waterways not covered by marine legislations. All states and territories apply this Transport Code through their regulatory system. For air transport, the Civil Aviation Act 1988 adopts the requirements of the ICAO Technical Instructions for the Safe Transport of Dangerous Goods by Air DOC 9284, which also adopts TS-R-1. The security of radioactive material in air transport is achieved via the Aviation Transport Security Act 2004. For sea transport Australian Marine Order 41 applies the requirements of IMDG (International Maritime Dangerous Goods) Code which also adopts TS-R-1. The security of radioactive material (nuclear material) is governed by two Commonwealth Agencies namely, ARPANSA and ASNO (Australian Safeguards and Non-proliferation Office) . ARPANSA regulates the security of radioactive sources through ARPANSA Code of Practice for the Security of Radioactive Sources 2007 which is based on the IAEA Draft Security Series. ASNO regulates security of nuclear material including U, Th and Pu through the Nuclear Non-Proliferation (Safeguards) Act, and the object of which is to give effect to certain obligations that Australia has as a party to the NPT, Australia’s safeguards agreement with the IAEA, and other bilateral safeguards agreements and certain obligations that Australia has as a party to the Convention for the Physical Protection of Nuclear Materials (CPPNM). This paper presents the effectiveness of regulatory approaches for safe and secure

Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material. Safety Guide (Spanish Edition)

International Nuclear Information System (INIS)

2010-01-01

This Safety Guide provides recommendations on achieving and demonstrating compliance with IAEA Safety Standards Series No. TS-R-1, Regulations for the Safe Transport of Radioactive Material, 2005 Edition, establishing safety requirements to be applied to the national and international transport of radioactive material. Transport is deemed to comprise all operations and conditions associated with and involved in the movement of radioactive material; these include the design, fabrication and maintenance of packaging, and the preparation, consigning, handling, carriage, storage in transit and receipt at the final destination of packages. This publication supersedes IAEA Safety Series No. TS-G-1.1, 2002 Edition

Modeling, analysis and comparison of TSR and OTC methods for MPPT and power smoothing in permanent magnet synchronous generator-based wind turbines

International Nuclear Information System (INIS)

Nasiri, M.; Milimonfared, J.; Fathi, S.H.

2014-01-01

Highlights: • Small signal modeling of PMSG wind turbine with two controllers are introduced. • Poles and zeroes analyzing of OTC and TSR methods is performed. • Generator output power with varying wind speed in PMSG wind turbine is studied. • MPPT capability of OTC and TSR methods to wind speed variations are compared. • Power smoothing capability and reducing mechanical stress of both methods are studied. - Abstract: This paper presents a small signal modeling of a direct-driven permanent magnet synchronous generator (PMSG) based on wind turbine which is connected to the grid via back-to-back converters. The proposed small signal model includes two maximum power point tracking (MPPT) controllers: tip speed ratio (TSR) control and optimal torque control (OTC). These methods are analytically compared to illustrate MPPT and power smoothing capability. Then, to compare the MPPT and power smoothing operation of the mentioned methods, simulations are performed in MATLAB/Simulink software. From the simulation results, OTC is highly efficient in power smoothing enhancement and has clearly good performance to extract maximum power from wind; however, TSR control has definitely fast responses to wind speed variations with the expense of higher fluctuations due to its non-minimum phase characteristic

Experimental investigation on thermochemical sulfate reduction in the presence of 1-pentanethiol at 200 and 250 °C: Implications for in situ TSR processes occurring in some MVT deposits

Science.gov (United States)

Yuan, Shunda; Ellis, Geoffrey S.; Chou, I-Ming; Burruss, Robert

2017-01-01

Organic sulfur compounds are ubiquitous in natural oil and gas fields and moderate-low temperature sulfide ore deposits. Previous studies have shown that organic sulfur compounds are important in enhancing the rates of thermochemical sulfate reduction (TSR) reactions, but the details of these reaction mechanisms remain unclear. In order to assess the extent of sulfate reduction in the presence of labile sulfur species at temperature conditions near to those where TSR occurs in nature, we conducted a series of experiments using the fused silica capillary capsule (FCSS) method. The tested systems containing labile sulfur species are MgSO4 + 1-pentanethiol (C5H11SH) + 1-octene (C8H16), MgSO4 + 1-octene (C8H16), MgSO4 + 1-pentanethiol (C5H11SH), 1-pentanethiol (C5H11SH)+H2O, and MgSO4 + 1-pentanethiol (C5H11SH) + ZnBr2 systems. Our results show that: (1) intermediate oxidized carbon species (ethanol and acetic acid) are formed during TSR simulation experiments when 1-pentanethiol is present; (2) in the presence of ZnBr2, 1-pentanethiol can be oxidized by sulfate to CO2 at 200 °C, which is within the temperature range observed in natural TSR; and (3) the precipitation of sulfide minerals may significantly promote the rate of TSR, indicating that the rates of in situ TSR reactions in ore deposits could be much faster than previously thought. This may be important for understanding the possibility of in situ TSR as a mechanism for the precipitation of metal sulfides in some ore deposits. These findings provide important experimental evidence for understanding the role of organic sulfur compounds in TSR reactions and the pathway of TSR reactions initiated by organic sulfur compounds under natural conditions.

Tie-down assessment of radioactive material packages on conveyances approach to competent authority approval applications in accordance with TS-R 1 (June 2000) and TS-G-1.1 (June 2002)

Energy Technology Data Exchange (ETDEWEB)

Read, J.; Owen, S.; Tso, C.F. [Arup, London (United Kingdom)

2004-07-01

This paper summarizes the authors' findings, during carrying out many package tie-down assessments, regarding the approach to the assessment process, and the issues that may arise. The current regulatory framework, as outlined in the abstract, is considered and possible areas for further development or research are discussed. The safe transport of a package containing radioactive material requires a secure system of retention on the conveyance during transport. The requirements for the assessment of this tie-down system are covered by International Atomic Energy Authority (IAEA) Regulations, and by the specific regulatory and guidance materials applying in the country of application. The IAEA regulations have recently been updated in TS-R-1 (June 2000) [1] and TS-G-1.1 (June 2002) [2]. In particular, Appendix V of reference [2] specifically covers 'Package Stowage and Retention during Transport'. In the United Kingdom the Transport Container Standardisation Committee (TCSC) provides guidance on the application of the IAEA regulations in TCSC 1006 ''The securing of Radioactive Materials on Conveyances'' [3], and this document has also been revised in December 2003. The basic requirement of the IAEA regulations is to maintain the integrity of the components of the package and its retention systems during routine operations (TS-R-1 Para 612). In accident conditions the package is permitted and may be required as part of the design (the weak-link approach as discussed later in this paper) to separate from the conveyance, while preserving the package integrity. Any tie-down attachments on the package shall be so designed that, under normal and accident conditions of transport, the forces in those attachments shall not impair the ability of the package to meet the requirements of the Regulations. (TS-R-1 Para 636).

Tie-down assessment of radioactive material packages on conveyances approach to competent authority approval applications in accordance with TS-R 1 (June 2000) and TS-G-1.1 (June 2002)

International Nuclear Information System (INIS)

Read, J.; Owen, S.; Tso, C.F.

2004-01-01

This paper summarizes the authors' findings, during carrying out many package tie-down assessments, regarding the approach to the assessment process, and the issues that may arise. The current regulatory framework, as outlined in the abstract, is considered and possible areas for further development or research are discussed. The safe transport of a package containing radioactive material requires a secure system of retention on the conveyance during transport. The requirements for the assessment of this tie-down system are covered by International Atomic Energy Authority (IAEA) Regulations, and by the specific regulatory and guidance materials applying in the country of application. The IAEA regulations have recently been updated in TS-R-1 (June 2000) [1] and TS-G-1.1 (June 2002) [2]. In particular, Appendix V of reference [2] specifically covers 'Package Stowage and Retention during Transport'. In the United Kingdom the Transport Container Standardisation Committee (TCSC) provides guidance on the application of the IAEA regulations in TCSC 1006 ''The securing of Radioactive Materials on Conveyances'' [3], and this document has also been revised in December 2003. The basic requirement of the IAEA regulations is to maintain the integrity of the components of the package and its retention systems during routine operations (TS-R-1 Para 612). In accident conditions the package is permitted and may be required as part of the design (the weak-link approach as discussed later in this paper) to separate from the conveyance, while preserving the package integrity. Any tie-down attachments on the package shall be so designed that, under normal and accident conditions of transport, the forces in those attachments shall not impair the ability of the package to meet the requirements of the Regulations. (TS-R-1 Para 636)

SAFETY EVALUATION OF OXALIC ACID WASTE RETRIEVAL IN SINGLE SHELL TANK (SST) 241-C-106

International Nuclear Information System (INIS)

SHULTZ, M.V.

2003-01-01

This report documents the safety evaluation of the process of retrieving sludge waste from single-shell tank 241-C-106 using oxalic acid. The results of the HAZOP, safety evaluation, and control allocation/decision are part of the report. This safety evaluation considers the use of oxalic acid to recover residual waste in single-shell tank (SST) 241-C-106. This is an activity not addressed in the current tank farm safety basis. This evaluation has five specific purposes: (1) Identifying the key configuration and operating assumptions needed to evaluate oxalic acid dissolution in SST 241-C-106. (2) Documenting the hazardous conditions identified during the oxalic acid dissolution hazard and operability study (HAZOP). (3) Documenting the comparison of the HAZOP results to the hazardous conditions and associated analyzed accident currently included in the safety basis, as documented in HNF-SD-WM-TI-764, Hazard Analysis Database Report. (4) Documenting the evaluation of the oxalic acid dissolution activity with respect to: (A) Accident analyses described in HNF-SD-WM-SAR-067, Tank Farms Final Safety Analysis Report (FSAR), and (B) Controls specified in HNF-SD-WM-TSR-006, Tank Farms Technical Safety Requirements (TSR). (5) Documenting the process and results of control decisions as well as the applicability of preventive and/or mitigative controls to each oxalic acid addition hazardous condition. This safety evaluation is not intended to be a request to authorize the activity. Authorization issues are addressed by the unreviewed safety question (USQ) evaluation process. This report constitutes an accident analysis

TSR: A storage and cooling ring for HIE-ISOLDE

Energy Technology Data Exchange (ETDEWEB)

Butler, P.A. [Oliver Lodge Laboratory, University of Liverpool, Liverpool L69 7ZE (United Kingdom); Blaum, K. [Max-Planck-Institut für Kernphysik, 69117 Heidelberg (Germany); Davinson, T. [School of Physics and Astronomy, University of Edinburgh, Edinburgh EH9 3JZ (United Kingdom); Flanagan, K.; Freeman, S.J. [School of Physics and Astronomy, University of Manchester, Manchester M13 9PL (United Kingdom); Grieser, M. [Max-Planck-Institut für Kernphysik, 69117 Heidelberg (Germany); Lazarus, I.H. [S.T.F.C. Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Litvinov, Yu.A. [Max-Planck-Institut für Kernphysik, 69117 Heidelberg (Germany); GSI Helmholtzzentrum für Schwerionenforschung, 64291 Darmstadt (Germany); Lotay, G. [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Page, R.D. [Oliver Lodge Laboratory, University of Liverpool, Liverpool L69 7ZE (United Kingdom); Raabe, R. [KU Leuven, Instituut voor Kern- en Stralingsfysica, 3001 Leuven (Belgium); Siesling, E.; Wenander, F. [CERN, 1211 Geneva 23 (Switzerland); Woods, P.J. [School of Physics and Astronomy, University of Edinburgh, Edinburgh EH9 3JZ (United Kingdom)

2016-06-01

It is planned to install the heavy-ion, low-energy ring TSR, currently at the Max-Planck-Institute for Nuclear Physics in Heidelberg, at the HIE-ISOLDE facility in CERN, Geneva. Such a facility will provide a capability for experiments with stored, cooled secondary beams that is rich and varied, spanning from studies of nuclear ground-state properties and reaction studies of astrophysical relevance, to investigations with highly-charged ions and pure isomeric beams. In addition to experiments performed using beams recirculating within the ring, the cooled beams can be extracted and exploited by external spectrometers for high-precision measurements. The capabilities of the ring facility as well as some physics cases will be presented, together with a brief report on the status of the project.

Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2003-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations; to be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources; and to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

Linking Safety Analysis to Safety Requirements

DEFF Research Database (Denmark)

Hansen, Kirsten Mark

Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...

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

Traceability of Software Safety Requirements in Legacy Safety Critical Systems

Science.gov (United States)

Hill, Janice L.

2007-01-01

How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

Safety of Nuclear Power Plants: Design. Specific Safety Requirements

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Structural and Thermal Safety Analysis Report for the Type B Radioactive Waste Transport Package

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Seo, K. S.; Lee, J. C.; Bang, K. S

2007-09-15

We carried out structural safety evaluation for the type B radioactive waste transport package. Requirements for type B packages according to the related regulations such as IAEA Safety Standard Series No. TS-R-1, Korea Most Act. 2001-23 and US 10 CFR Part 71 were evaluated. General requirements for packages such as those for a lifting attachment, a tie-down attachment and pressure condition were considered. For the type B radioactive waste transport package, the structural, thermal and containment analyses were carried out under the normal transport conditions. Also the safety analysis were conducted under the accidental transport conditions. The 9 m drop test, 1 m puncture test, fire test and water immersion test under the accidental transport conditions were consecutively done. The type B radioactive waste transport packages were maintained the structural and thermal integrities.

Safety of Research Reactors. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2017-01-01

This Safety Requirements publication establishes requirements for all main areas of safety for research reactors, with particular emphasis on requirements for design and operation. It explains the safety objectives and concepts that form the basis for safety and safety assessment for all stages in the lifetime of a research reactor. Technical and administrative requirements for the safety of new research reactors are established in accordance with these objectives and concepts, and they are to be applied to the extent practicable for existing research reactors. The safety requirements established in this publication for the management of safety and regulatory supervision apply to site evaluation, design, manufacturing, construction, commissioning, operation (including utilization and modification), and planning for decommissioning of research reactors (including critical assemblies and subcritical assemblies). The publication is intended for use by regulatory bodies and other organizations with responsibilities in these areas and in safety analysis, verification and review, and the provision of technical support.

Safety of nuclear fuel cycle facilities. Safety requirements

International Nuclear Information System (INIS)

2008-01-01

This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

Leadership and Management for Safety. General Safety Requirements

International Nuclear Information System (INIS)

2016-01-01

This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factor, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations (registrants and licensees) and other organizations concerned with facilities and activities that give rise to radiation risks

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

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

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

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

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Schedules of Provisions of the IAEA Regulations for the Safe Transport of Radioactive Material (2009 Ed.). Safety Guide

International Nuclear Information System (INIS)

2014-01-01

This Safety Guide is issued in support of Regulations for the Safe Transport of Radioactive Material (IAEA Safety Standards Series No. TS-R-1, 2009 Edition). It lists the paragraph numbers of the Transport Regulations that are relevant for specified types of consignment, classified according to their UN numbers. It does not provide additional recommendations. The intended users are consignors and consignees, carriers, shippers, regulators, and end users involved in the transport of radioactive material. A person or organization intending to transport a particular type of consignment of radioactive material must meet requirements in all sections of the Transport Regulations. This Safety Guide aids users by providing a listing of the relevant requirements of the Transport Regulations for each type of radioactive material, package or shipment. Once a consignor has classified the radioactive material to be shipped, the appropriate UN number can be assigned and the paragraph numbers of the requirements that apply for the shipment can be found in the corresponding schedule

Supplement to safety analysis report. 306-W building operations safety requirement

International Nuclear Information System (INIS)

Richey, C.R.

1979-08-01

The operations safety requirements (OSRs) presented in this report define the conditions, safe boundaries, and management control needed for safely conducting operations with radioactive materials in the Pacific Northwest Laboratory (PNL) 306-W building. The safety requirements are organized in five sections. Safety limits are safety-related process variables that are observable and measurable. Limiting conditions cover: equipment and technical conditions and characteristics of the facility and operations necessary for continued safe operation. Surveillance requirements prescribe the requirements for checking systems and components that are essential to safety. Equipment design controls require that changes to process equipment and systems be independently checked and approved to assure that the changes will have no adverse effect on safety. Administrative controls describe and discuss the organization and administrative systems and procedures to be used for safe operation of the facility. Details of the implementation of the operations safety requirements are prescribed by internal PNL documents such as criticality safety specifications and radiation work procedures

Yucca Mountain Project technical status report (TSR), October 1989--March 1990

International Nuclear Information System (INIS)

1990-01-01

This Yucca Mountain Project Technical Status Report (TSR) on site characterization is the second in a series of reports that will be issued at approximately six-month intervals during site characterization. In addition, progress made toward the initiation and conduct of new site characterization activities is included. For this report, information on the technical progress made by Yucca Mountain Project participating organizations has been compiled covering the period from October 1989, through March 31, 1990. The status report consists of three sections: an introductory section; a section on the status of site characterization, which includes preparatory activities, sites programs, repository design, seals system design, waste package design, and performance assessment; and a reference section, which provides a complete listing of all published documents cited in the text. 59 refs

Safety of magnetic fusion facilities: Requirements

International Nuclear Information System (INIS)

1996-05-01

This Standard identifies safety requirements for magnetic fusion facilities. Safety functions are used to define outcomes that must be achieved to ensure that exposures to radiation, hazardous materials, or other hazards are maintained within acceptable limits. Requirements applicable to magnetic fusion facilities have been derived from Federal law, policy, and other documents. In addition to specific safety requirements, broad direction is given in the form of safety principles that are to be implemented and within which safety can be achieved

IAEA safety requirements for safety assessment of fuel cycle facilities and activities

International Nuclear Information System (INIS)

Jones, G.

2013-01-01

The IAEA's Statute authorises the Agency to establish standards of safety for protection of health and minimisation of danger to life and property. In that respect, the IAEA has established a Safety Fundamentals publication which contains ten safety principles for ensuring the protection of workers, the public and the environment from the harmful effects of ionising radiation. A number of these principles require safety assessments to be carried out as a means of evaluating compliance with safety requirements for all nuclear facilities and activities and to determine the measures that need to be taken to ensure safety. The safety assessments are required to be carried out and documented by the organisation responsible for operating the facility or conducting the activity, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorisation process. In addition to the principles of the Safety Fundamentals, the IAEA establishes requirements that must be met to ensure the protection of people and the environment and which are governed by the principles in the Safety Fundamentals. The IAEA's Safety Requirements publication 'Safety Assessment for Facilities and Activities', establishes the safety requirements that need to be fulfilled in conducting and maintaining safety assessments for the lifetime of facilities and activities, with specific attention to defence in depth and the requirement for a graded approach to the application of these safety requirements across the wide range of fuel cycle facilities and activities. Requirements for independent verification of the safety assessment that needs to be carried out by the operating organisation, including the requirement for the safety assessment to be periodically reviewed and updated are also covered. For many fuel cycle facilities and activities, environmental impact assessments and non-radiological risk assessments will be required. The

TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS

International Nuclear Information System (INIS)

KRIPPS, L.J.

2006-01-01

This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for natural event hazard (NEH)-initiated accidents. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls

2011 Annual Criticality Safety Program Performance Summary

Energy Technology Data Exchange (ETDEWEB)

Andrea Hoffman

2011-12-01

The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The

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

International Nuclear Information System (INIS)

2012-01-01

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

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

International Nuclear Information System (INIS)

2012-01-01

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

MODEL 9977 B(M)F-96 SAFETY ANALYSIS REPORT FOR PACKAGING

Energy Technology Data Exchange (ETDEWEB)

Abramczyk, G; Paul Blanton, P; Kurt Eberl, K

2006-05-18

This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on and for the 9977 Shipping Package, referred to as the General Purpose Fissile Package (GPFP). The performance evaluation presented in this SARP documents the compliance of the 9977 package with the regulatory safety requirements for Type B packages. Per 10 CFR 71.59, for the 9977 packages evaluated in this SARP, the value of ''N'' is 50, and the Transport Index based on nuclear criticality control is 1.0. The 9977 package is designed with a high degree of single containment. The 9977 complies with 10 CFR 71 (2002), Department of Energy (DOE) Order 460.1B, DOE Order 460.2, and 10 CFR 20 (2003) for As Low As Reasonably Achievable (ALARA) principles. The 9977 also satisfies the requirements of the Regulations for the Safe Transport of Radioactive Material--1996 Edition (Revised)--Requirements. IAEA Safety Standards, Safety Series No. TS-R-1 (ST-1, Rev.), International Atomic Energy Agency, Vienna, Austria (2000). The 9977 package is designed, analyzed and fabricated in accordance with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 1992 edition.

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)

ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

WILLIAMS, J.C.

2003-11-15

This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

International Nuclear Information System (INIS)

2016-01-01

This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication

Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

International Nuclear Information System (INIS)

2017-01-01

This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

Site evaluation for nuclear installations. Safety requirements

International Nuclear Information System (INIS)

2003-01-01

This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Siting, which was issued in 1988 as Safety Series No. 50-C-S (Rev. 1). It takes account of developments relating to site evaluations for nuclear installations since the Code on Siting was last revised. These developments include the issuing of the Safety Fundamentals publication on The Safety of Nuclear Installations, and the revision of various safety standards and other publications relating to safety. Requirements for site evaluation are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear installations. It is recognized that there are steady advances in technology and scientific knowledge, in nuclear safety and in what is considered adequate protection. Safety requirements change with these advances and this publication reflects the present consensus among States. This Safety Requirements publication was prepared under the IAEA programme on safety standards for nuclear installations. It establishes requirements and provides criteria for ensuring safety in site evaluation for nuclear installations. The Safety Guides on site evaluation listed in the references provide recommendations on how to meet the requirements established in this Safety Requirements publication. The objective of this publication is to establish the requirements for the elements of a site evaluation for a nuclear installation so as to characterize fully the site specific conditions pertinent to the safety of a nuclear installation. The purpose is to establish requirements for criteria, to be applied as appropriate to site and site-installation interaction in operational states and accident conditions, including those that could lead to emergency measures for: (a) Defining the extent of information on a proposed site to be presented by the applicant; (b) Evaluating a proposed site to ensure that the site

Range Flight Safety Requirements

Science.gov (United States)

Loftin, Charles E.; Hudson, Sandra M.

2018-01-01

The purpose of this NASA Technical Standard is to provide the technical requirements for the NPR 8715.5, Range Flight Safety Program, in regards to protection of the public, the NASA workforce, and property as it pertains to risk analysis, Flight Safety Systems (FSS), and range flight operations. This standard is approved for use by NASA Headquarters and NASA Centers, including Component Facilities and Technical and Service Support Centers, and may be cited in contract, program, and other Agency documents as a technical requirement. This standard may also apply to the Jet Propulsion Laboratory or to other contractors, grant recipients, or parties to agreements to the extent specified or referenced in their contracts, grants, or agreements, when these organizations conduct or participate in missions that involve range flight operations as defined by NPR 8715.5.1.2.2 In this standard, all mandatory actions (i.e., requirements) are denoted by statements containing the term “shall.”1.3 TailoringTailoring of this standard for application to a specific program or project shall be formally documented as part of program or project requirements and approved by the responsible Technical Authority in accordance with NPR 8715.3, NASA General Safety Program Requirements.

Disposal of Radioactive Waste. Specific Safety Requirements

International Nuclear Information System (INIS)

2011-01-01

This publication establishes requirements applicable to all types of radioactive waste disposal facility. It is linked to the fundamental safety principles for each disposal option and establishes a set of strategic requirements that must be in place before facilities are developed. Consideration is also given to the safety of existing facilities developed prior to the establishment of present day standards. The requirements will be complemented by Safety Guides that will provide guidance on good practice for meeting the requirements for different types of waste disposal facility. Contents: 1. Introduction; 2. Protection of people and the environment; 3. Safety requirements for planning for the disposal of radioactive waste; 4. Requirements for the development, operation and closure of a disposal facility; 5. Assurance of safety; 6. Existing disposal facilities; Appendices.

Leadership and Management for Safety. General Safety Requirements (Arabic Edition)

International Nuclear Information System (INIS)

2016-01-01

This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

Leadership and Management for Safety. General Safety Requirements (Chinese Edition)

International Nuclear Information System (INIS)

2016-01-01

This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

Leadership and Management for Safety. General Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2016-01-01

This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

Leadership and Management for Safety. General Safety Requirements (Spanish Edition)

International Nuclear Information System (INIS)

2017-01-01

his Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

Drug safety alerts of pharmacovigilance programme of India: A scope for targeted spontaneous reporting in India

Directory of Open Access Journals (Sweden)

Prasad Thota

2018-01-01

Conclusion: In India, spontaneous reporting of ADRs existed since 1998 under passive surveillance method, but there is an urgent need to initiate TSR, which is a complementary method to spontaneous reporting on these drug safety alerts for further regulatory action by Central Drugs Standard Control Organization.

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

International Nuclear Information System (INIS)

2017-01-01

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

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

International Nuclear Information System (INIS)

2016-01-01

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

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

International Nuclear Information System (INIS)

2017-01-01

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

Laser spectroscopy with a cooler ring at the ESR (GSI) and the TSR (MPI Heidelberg)

International Nuclear Information System (INIS)

Kuehl, T.; Borneis, S.; Greten, G.; Marx, D.; Neumann, R.; Schroeder, S.; Grieser, R.; Hoog, I.; Huber, G.; Klaft, I.; Klein, R.; Merz, P.; Balykin, V.; Bock, M.; Ellert, C.; Forck, P.; Grieser, M.; Grimm, R.; Habs, D.; Miesner, H.J.; Petrich, W.; Wanner, B.; Becker, C.; Schwalm, D.; Wolf, A.

1992-01-01

At the TSR cooler ring at Heidelberg, laser studies were carried out using singly charged lithium and beryllium ions. Laser spectroscopy of relativistic lithium ions (υ = 0.04c) yielded signals with a narrow linewidth, suitable for an experimental test of special relativity. A dramatic reduction of the beam temperature, as defined by the longitudinal velocity spread, was achieved via laser cooling in both cases. At the ion energies available at ESR it will become possible to prepare and store bare ions up to U 92+ . Electron cooling was successfully demonstrated for hydrogen-like Bi 82+ ions, where a laser experiment is scheduled to study the ground-state hyperfine splitting. (orig.)

The development of safety requirements

International Nuclear Information System (INIS)

Jorel, M.

2009-01-01

This document describes the safety approach followed in France for the design of nuclear reactors. This safety approach is based on safety principles from which stem safety requirements that set limiting values for specific parameters. The improvements in computerized simulation, the use of more adequate new materials, a better knowledge of the concerned physical processes, the changes in the reactor operations (higher discharge burnups for instance) have to be taken into account for the definition of safety criteria and the setting of limiting values. The developments of the safety criteria linked to the risks of cladding failure and loss of primary coolant are presented. (A.C.)

Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

Safety Design Approach for the Development of Safety Requirements for Design of Commercial HTGR

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, Xing; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-01-01

The research committee on “Safety requirements for HTGR design” was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactors (HTGRs), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGRs which is a basement of the safety requirements is determined prior to the development of the safety requirements. The safety design approaches for the commercial HTGRs are to confine the radioactive materials within the coated fuel particles not only during normal operation but also during accident conditions, and the integrity of the coated fuel particles and other requiring physical barriers are protected by the inherent and passive safety features. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGRs determined in the research committee. (author)

Safety assessment for facilities and activities. General safety requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF 6 ; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2010-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation.? read more The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are

Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (Arabic Edition)

International Nuclear Information System (INIS)

2017-01-01

This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

Radiation safety requirements for radionuclide laboratories

International Nuclear Information System (INIS)

1993-01-01

In accordance with the section 26 of the Finnish Radiation Act (592/91) the safety requirements to be taken into account in planning laboratories and other premises, which affect safety in the use of radioactive materials, are confirmed by the Finnish Centre for Radiation and Nuclear Safety. The guide specifies the requirements for laboratories and storage rooms in which radioactive materials are used or stored as unsealed sources. There are also some general instructions concerning work procedures in a radionuclide laboratory

Investigation on regulatory requirements for radiation safety management

International Nuclear Information System (INIS)

Han, Eun Ok; Choi, Yoon Seok; Cho, Dae Hyung

2013-01-01

NRC recognizes that efficient management of radiation safety plan is an important factor to achieve radiation safety service. In case of Korea, the contents to perform the actual radiation safety management are legally contained in radiation safety management reports based on the Nuclear Safety Act. It is to prioritize the importance of safety regulations in each sector in accordance with the current situation of radiation and radioactive isotopes-used industry and to provide a basis for deriving safety requirements and safety regulations system maintenance by the priority of radiation safety management regulations. It would be helpful to achieve regulations to conform to reality based on international standards if consistent safety requirements is developed for domestic users, national standards and international standards on the basis of the results of questions answered by radiation safety managers, who lead on-site radiation safety management, about the priority of important factors in radioactive sources use, sales, production, moving user companies, to check whether derived configuration requirements for radiation safety management are suitable for domestic status

Long-time dynamics of laser-cooled ions in the TSR storage ring

International Nuclear Information System (INIS)

Mudrich, M.

2000-01-01

This diploma thesis studies experimentally the long-time dynamics of laser-cooled 9 Be + -beams at the TSR under different cooling conditions. The goal is to enlarge the understanding of ultra-cold, non-neutral plasma at high center-of-mass energies. By means of improved measurement capabilities one can now for the first time monitor the entire phase-space over a long time. This makes it possible to quantitatively analyse the possibilities and limitations of laser cooling at a storage ring. Under optimum cooling conditions a regime of high phase-space density is reached, close to the region where influences of Coulomb coupling are expected. Furthermore, a Monte-Carlo model is worked out that qualitatively describes the beam dynamics. The model includes the influence of transverse-longitudinal coupling due to intra beam scattering on the longitudinal phase-space distribution. At high phase-space density a sudden disappearance of intra beam collisions is observed experimentally and possible interpretations are given. (orig.)

Control Decisions for Flammable Gas Hazards in Double Contained Receiver Tanks (DCRTs)

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2000-06-28

This report describes the control decisions for flammable gas hazards in double-contained receiver tanks (DCRTs) made at control decision meetings on November 16, 17, and 18, 1999, on April 19,2000, and on May 10,2000, and their basis. These control decisions, and the analyses that support them, will be documented in an amendment to the Final Safety Analysis Report (FSAR) (CHG 2000a) and Technical Safety Requirements (TSR) (CHG 2000b) to close the Flammable Gas Unreviewed Safety Question (USQ) (Bacon 1996 and Wagoner 1996) for DCRTs. Following the contractor Tier I review of the FSAR and TSR amendment, it will be submitted to the U.S. Department of Energy (DOE), Office of River Protection (ORP) for review and approval.

Technical Safety Requirements for the B695 Segment of the Decontamination and Waste Treatment Facility

International Nuclear Information System (INIS)

Larson, H L

2007-01-01

waste. Upon approval of the permit modification, B696S rooms 1007, 1008, and 1009 will be able to store hazardous and mixed waste for up to 1 year. Furthermore, an additional drum crusher and a Waste Packaging Unit will be permitted to treat hazardous and mixed waste. RHWM generally processes LLW with no, or extremely low, concentrations of transuranics (i.e., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., 90 Sr, 137 Cs, 3 H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment of the DWTF. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment of the DWTF. A basis explanation follows each of the requirements described in Section 5.5, Specific Administrative Controls. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6

Safety requirements applicable to the SMART design

International Nuclear Information System (INIS)

Seul, Kwang Won; Kim, Wee Kyong; Kim, Hho Jung

1999-01-01

The 330 MW thermal power of integral reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at KAERI for seawater desalination application and electricity generation. The final product of nuclear desalination plant (NDP) is electricity and fresh water. Thus, in addition to the protection of the public around the plant facility from the possible release of radioactive materials, the fresh water should be prevented from radioactivity contamination. In this study, to ensure the safety of SMART reactor in the early stage of design development, the safety requirements applicable to the SMART design were investigated, based on the current regulatory requirements for the existing NPPs and the advanced light water reactor (LWR) designs. The interface requirements related to the desalination facility were also investigated, based on the recent IAEA research activities pertaining to the NDP. As a result, it was found that the current regulatory requirements and guidance for the existing NPPs and advanced LWR designs are applicable to the SMART design and its safety evaluation. However, the safety requirements related to the SMART-specific design and the desalination plant are needed to develop in the future to assure the safety of the SMART reactor

HTR-PM Safety requirement and Licensing experience

International Nuclear Information System (INIS)

Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

2014-01-01

HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

Technical basis document for natural event hazards

International Nuclear Information System (INIS)

CARSON, D.M.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for natural event hazards (NEH)-initiated representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

Geological disposal of radioactive waste. Safety requirements

International Nuclear Information System (INIS)

2006-01-01

This Safety Requirements publication is concerned with providing protection to people and the environment from the hazards associated with waste management activities related to disposal, i.e. hazards that could arise during the operating period and following closure. It sets out the protection objectives and criteria for geological disposal and establishes the requirements that must be met to ensure the safety of this disposal option, consistent with the established principles of safety for radioactive waste management. It is intended for use by those involved in radioactive waste management and in making decisions in relation to the development, operation and closure of geological disposal facilities, especially those concerned with the related regulatory aspects. This publication contains 1. Introduction; 2. Protection of human health and the environment; 3. The safety requirements for geological disposal; 4. Requirements for the development, operation and closure of geological disposal facilities; Appendix: Assurance of compliance with the safety objective and criteria; Annex I: Geological disposal and the principles of radioactive waste management; Annex II: Principles of radioactive waste management

Site safety requirements for high level waste disposal

International Nuclear Information System (INIS)

Chen Weiming; Wang Ju

2006-01-01

This paper outlines the content, status and trend of site safety requirements of International Atomic Energy Agency, America, France, Sweden, Finland and Japan. Site safety requirements are usually represented as advantageous vis-a-vis disadvantagous conditions, and potential advantage vis-a-vis disadvantage conditions, respectively in aspects of geohydrology, geochemistry, lithology, climate and human intrusion etc. Study framework and steps of site safety requirements for China are discussed under the view of systems science. (authors)

New requirements on safety of nuclear power plants according to the IAEA safety standards

International Nuclear Information System (INIS)

Misak, J.

2005-01-01

In this presentation author presents new requirements on safety of nuclear power plants according to the IAEA safety standards. It is concluded that: - New set of IAEA Safety Standards is close to completion: around 40 standards for NPPs; - Different interpretation of IAEA Safety Standards at present: best world practices instead of previous 'minimum common denominator'; - A number of safety improvements required for NPPs; - Requirements related to BDBAs and severe accidents are the most demanding due to degradation of barriers: hardware modifications and accident management; - Large variety between countries in implementation of accident management programmes: from minimum to major hardware modifications; -Distinction between existing and new NPPs is essential from the point of view of the requirements; WWER 440 reactors have potential to reflect IAEA Safety Standards for existing NPPs; relatively low reactor power offers broader possibilities

NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

International Nuclear Information System (INIS)

Marshall, A.C.; Lee, J.H.; McCulloch, W.H.; Sawyer, J.C. Jr.; Bari, R.A.; Brown, N.W.; Cullingford, H.S.; Hardy, A.C.; Remp, K.; Sholtis, J.A.

1992-01-01

An Interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) 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 requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed

Mixing of incompatible materials in waste tanks technical basis document

International Nuclear Information System (INIS)

SANDGREN, K.R.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the FR-equency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

Technical Safety Requirements for the B695 Segment of the Decontamination and Waste Treatment Facility

Energy Technology Data Exchange (ETDEWEB)

Larson, H L

2007-09-07

mixed waste. Upon approval of the permit modification, B696S rooms 1007, 1008, and 1009 will be able to store hazardous and mixed waste for up to 1 year. Furthermore, an additional drum crusher and a Waste Packaging Unit will be permitted to treat hazardous and mixed waste. RHWM generally processes LLW with no, or extremely low, concentrations of transuranics (i.e., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, {sup 3}H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment of the DWTF. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment of the DWTF. A basis explanation follows each of the requirements described in Section 5.5, Specific Administrative Controls. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6.

Analyzing Software Requirements Errors in Safety-Critical, Embedded Systems

Science.gov (United States)

Lutz, Robyn R.

1993-01-01

This paper analyzes the root causes of safety-related software errors in safety-critical, embedded systems. The results show that software errors identified as potentially hazardous to the system tend to be produced by different error mechanisms than non- safety-related software errors. Safety-related software errors are shown to arise most commonly from (1) discrepancies between the documented requirements specifications and the requirements needed for correct functioning of the system and (2) misunderstandings of the software's interface with the rest of the system. The paper uses these results to identify methods by which requirements errors can be prevented. The goal is to reduce safety-related software errors and to enhance the safety of complex, embedded systems.

Meeting the maglev system's safety requirements

Energy Technology Data Exchange (ETDEWEB)

Pierick, K

1983-12-01

The author shows how the safety requirements of the maglev track system derive from the general legal conditions for the safety of tracked transport. It is described how their compliance beyond the so-called ''development-accompanying'' and ''acceptance-preparatory'' safety work can be assured for the Transrapid test layout (TVE) now building in Emsland and also for later application as public transport system in Germany within the meaning of the General Railway Act.

Generic Safety Requirements for Developing Safe Insulin Pump Software

Science.gov (United States)

Zhang, Yi; Jetley, Raoul; Jones, Paul L; Ray, Arnab

2011-01-01

Background The authors previously introduced a highly abstract generic insulin infusion pump (GIIP) model that identified common features and hazards shared by most insulin pumps on the market. The aim of this article is to extend our previous work on the GIIP model by articulating safety requirements that address the identified GIIP hazards. These safety requirements can be validated by manufacturers, and may ultimately serve as a safety reference for insulin pump software. Together, these two publications can serve as a basis for discussing insulin pump safety in the diabetes community. Methods In our previous work, we established a generic insulin pump architecture that abstracts functions common to many insulin pumps currently on the market and near-future pump designs. We then carried out a preliminary hazard analysis based on this architecture that included consultations with many domain experts. Further consultation with domain experts resulted in the safety requirements used in the modeling work presented in this article. Results Generic safety requirements for the GIIP model are presented, as appropriate, in parameterized format to accommodate clinical practices or specific insulin pump criteria important to safe device performance. Conclusions We believe that there is considerable value in having the diabetes, academic, and manufacturing communities consider and discuss these generic safety requirements. We hope that the communities will extend and revise them, make them more representative and comprehensive, experiment with them, and use them as a means for assessing the safety of insulin pump software designs. One potential use of these requirements is to integrate them into model-based engineering (MBE) software development methods. We believe, based on our experiences, that implementing safety requirements using MBE methods holds promise in reducing design/implementation flaws in insulin pump development and evolutionary processes, therefore improving

Cold Vacuum Drying (CVD) Facility Technical Safety Requirements

International Nuclear Information System (INIS)

KRAHN, D.E.

2000-01-01

The Technical Safety Requirements (TSRs) for the Cold Vacuum Drying Facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt of multi-canister overpacks (MCOs) containing spent nuclear fuel. removal of free water from the MCOs using the cold vacuum drying process, and inerting and testing of the MCOs before transport to the Canister Storage Building. Controls required for public safety, significant defense in depth, significant worker safety, and for maintaining radiological and toxicological consequences below risk evaluation guidelines are included

78 FR 46560 - Pipeline Safety: Class Location Requirements

Science.gov (United States)

2013-08-01

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part... class location requirements for gas transmission pipelines. Section 5 of the Pipeline Safety, Regulatory... and, with respect to gas transmission pipeline facilities, whether applying IMP requirements to...

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

International Nuclear Information System (INIS)

MORENO, M.R.

2004-01-01

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

Operating manual for the Tower Shielding Facility

International Nuclear Information System (INIS)

1985-12-01

This manual provides information necessary to operate and perform maintenance on the reactor systems and all equipment or systems which can affect their operation or the safety of personnel at the Tower Shielding Facility. The first four chapters consist of introductory and descriptive material of benefit to personnel in training, the qualifications required for training, the responsibilities of the personnel in the organization, and the procedures for reviewing proposed experiments. Chapter 8, Emergency Procedures, is also a necessary part of the indoctrination of personnel. The procedures for operation of the Tower Shielding Reactor (TSR-II), its water cooling system, and the main tower hoists are outlined in Chapters 5, 6, and 7. The Technical Specification surveillance requirements for the TSR-II are summarized in Chapter 9. The maintenance and calibration schedule is spelled out in Chapter 10. The procedures for assembly and disassembly of the TSR-II are outlined in Chapter 11

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

International Nuclear Information System (INIS)

2009-01-01

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

International standardization of safety requirements for fast reactors

International Nuclear Information System (INIS)

2011-06-01

Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

Safety design guides for seismic requirements for CANDU 9

International Nuclear Information System (INIS)

Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

1996-03-01

This safety design guide for seismic requirements for CANDU 9 describes the seismic design philosophy, defines the applicable earthquakes and identifies the structures and systems requiring seismic qualification to ensure that the essential safety function can be adequately satisfied following earthquake. The detailed requirements for structures, systems and components which must be seismically qualified are specified in the Appendix. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 fig., (Author) .new

Disposal of Radioactive Waste. Specific Safety Requirements (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This Safety Requirements publication applies to the disposal of radioactive waste of all types by means of emplacement in designed disposal facilities, subject to the necessary limitations and controls being placed on the disposal of the waste and on the development, operation and closure of facilities. The classification of radioactive waste is discussed. This Safety Requirements publication establishes requirements to provide assurance of the radiation safety of the disposal of radioactive waste, in the operation of a disposal facility and especially after its closure. The fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. This is achieved by setting requirements on the site selection and evaluation and design of a disposal facility, and on its construction, operation and closure, including organizational and regulatory requirements.

Fire safety requirements for electrical cables towards nuclear reactor safety

International Nuclear Information System (INIS)

Raju, M.R.

2002-01-01

Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

The main requirements of the International Basic Safety Standards

International Nuclear Information System (INIS)

Webb, G.A.M.

1998-01-01

The main requirements of the new international basic safety standards are discussed, including such topics as health effects of ionizing radiations, the revision of basic safety standards, the requirements for radiation protection practices, the requirements for intervention,and the field of regulatory infrastructures. (A.K.)

Safety Requirements and Modern Technical Requirements in Human Information Systems in Amman Hotels

OpenAIRE

Farouq Ahmad Alazzam; Sattam Rakan Allahawiah; Mohammad Nayef Alsarayreh; Kafa Hmoud Abdallah al Nawaiseh

2015-01-01

This study aimed to demonstrate the availability of Safety requirements and modern technical requirements in human information systems in Amman hotels. an the most important results of this study is the availability of security and safety requirements in human information systems In Amman hotels and The adequacy of the information that it provided .and show that all departments are not connected by appropriate and effective communication networks in adequate form . Also sophisticated operatin...

Waste Encapsulation and Storage Facility interim operational safety requirements

CERN Document Server

Covey, L I

2000-01-01

The Interim Operational Safety Requirements (IOSRs) for the Waste Encapsulation and Storage Facility (WESF) define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt and inspection of cesium and strontium capsules from private irradiators; decontamination of the capsules and equipment; surveillance of the stored capsules; and maintenance activities. Controls required for public safety, significant defense-in-depth, significant worker safety, and for maintaining radiological consequences below risk evaluation guidelines (EGs) are included.

OSHA safety requirements for hazardous chemicals in the workplace.

Science.gov (United States)

Dohms, J

1992-01-01

This article outlines the Occupational Safety and Health Administration (OSHA) requirements set forth by the Hazard Communication Standard, which has been in effect for the healthcare industry since 1987. Administrators who have not taken concrete steps to address employee health and safety issues relating to hazardous chemicals are encouraged to do so to avoid the potential of large fines for cited violations. While some states administer their own occupational safety and health programs, they must adopt standards and enforce requirements that are at least as effective as federal requirements.

The Canadian Nuclear Safety Commission's financial guarantee requirements

International Nuclear Information System (INIS)

Ferch, R.

2006-01-01

The Nuclear Safety and Control Act gives the Canadian Nuclear Safety Commission (CNSC) the legal authority to require licensees to provide financial guarantees in order to meet the purposes of the Act. CNSC policy and guidance with regard to financial guarantees is outlined, and the current status of financial guarantee requirements as applied to various CNSC licensees is described. (author)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-04-15

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

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

International Nuclear Information System (INIS)

1991-12-01

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

High-Speed Maglev Trains; German Safety Requirements

Science.gov (United States)

1991-12-31

This document is a translation of technology-specific safety requirements developed : for the German Transrapid Maglev technology. These requirements were developed by a : working group composed of representatives of German Federal Railways (DB), Tes...

Development of High-Level Safety Requirements for a Pyroprocessing Facility

Energy Technology Data Exchange (ETDEWEB)

Seo, Seok Jun; Jo, Woo Jin; You, Gil Sung; Choung, Won Myung; Lee, Ho Hee; Kim, Hyun Min; Jeon, Hong Rae; Ku, Jeong Hoe; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Korea Atomic Energy Research Institute (KAERI) has been developing a pyroproceesing technology to reduce the waste volume and recycle some elements. The pyroprocessing includes several treatment processes which are related with not only radiological and physical but also chemical and electrochemical properties. Thus, it is of importance to establish safety design requirements considering all the aspects of those properties for a reliable pyroprocessing facility. In this study, high-level requirements are presented in terms of not only radiation protection, nuclear criticality, fire protection, and seismic safety but also confinement and chemical safety for the unique characteristics of a pyroprocessing facility. Several high-level safety design requirements such as radiation protection, nuclear criticality, fire protection, seismic, confinement, and chemical processing were presented for a pyroprocessing facility. The requirements must fulfill domestic and international safety technology standards for a nuclear facility. Furthermore, additional requirements should be considered for the unique electrochemical treatments in a pyroprocessing facility.

Discussion on several important safety requirements for the new nuclear power plant

International Nuclear Information System (INIS)

Yan Tianwen; Li Jigen; Zhang Lin; Feng Youcai; Jia Xiang; Li Wenhong

2013-01-01

Post the Fukushima nuclear accident, the Chinese government raised higher safety goals and safety requirements for the new nuclear power plant to be constructed. The paper expounded the important indicators of safety requirements and the aspects of safety modification that had been developed for the new NPPs. It also discussed and analyzed the main fields required by the new NPPs safety requirements in the safety goals, safety evaluation of sites, defenses of internal and external events, severe accident prevention and mitigation, design of reactor core, containment system and I and C system, and optimization of engineering measure, which gave some references to the design, construction and safety modifications of new NPPs in China. (authors)

Westinghouse Hanford Company safety analysis reports and technical safety requirements upgrade program

International Nuclear Information System (INIS)

Busche, D.M.

1995-09-01

During Fiscal Year 1992, the US Department of Energy, Richland Operations Office (RL) separately transmitted the following US Department of Energy (DOE) Orders to Westinghouse Hanford Company (WHC) for compliance: DOE 5480.21, ''Unreviewed Safety Questions,'' DOE 5480.22, ''Technical Safety Requirements,'' and DOE 5480.23, ''Nuclear Safety Analysis Reports.'' WHC has proceeded with its impact assessment and implementation process for the Orders. The Orders are closely-related and contain some requirements that are either identical, similar, or logically-related. Consequently, WHC has developed a strategy calling for an integrated implementation of the three Orders. The strategy is comprised of three primary objectives, namely: Obtain DOE approval of a single list of DOE-owned and WHC-managed Nuclear Facilities, Establish and/or upgrade the ''Safety Basis'' for each Nuclear Facility, and Establish a functional Unreviewed Safety Question (USQ) process to govern the management and preservation of the Safety Basis for each Nuclear Facility. WHC has developed policy-revision and facility-specific implementation plans to accomplish near-term tasks associated with the above strategic objectives. This plan, which as originally submitted in August 1993 and approved, provided an interpretation of the new DOE Nuclear Facility definition and an initial list of WHC-managed Nuclear Facilities. For each current existing Nuclear Facility, existing Safety Basis documents are identified and the plan/status is provided for the ISB. Plans for upgrading SARs and developing TSRs will be provided after issuance of the corresponding Rules

A TSR Visual Servoing System Based on a Novel Dynamic Template Matching Method

Directory of Open Access Journals (Sweden)

Jia Cai

2015-12-01

Full Text Available The so-called Tethered Space Robot (TSR is a novel active space debris removal system. To solve its problem of non-cooperative target recognition during short-distance rendezvous events, this paper presents a framework for a real-time visual servoing system using non-calibrated monocular-CMOS (Complementary Metal Oxide Semiconductor. When a small template is used for matching with a large scene, it always leads to mismatches, so a novel template matching algorithm to solve the problem is presented. Firstly, the novel matching algorithm uses a hollow annulus structure according to a FAST (Features from Accelerated Segment algorithm and makes the method be rotation-invariant. Furthermore, the accumulative deviation can be decreased by the hollow structure. The matching function is composed of grey and gradient differences between template and object image, which help it reduce the effects of illumination and noises. Then, a dynamic template update strategy is designed to avoid tracking failures brought about by wrong matching or occlusion. Finally, the system synthesizes the least square integrated predictor, realizing tracking online in complex circumstances. The results of ground experiments show that the proposed algorithm can decrease the need for sophisticated computation and improves matching accuracy.

Development of a user interface for the TSR and its application for beam diagnostics

International Nuclear Information System (INIS)

Tetzlaff, K.

1997-01-01

One of the topics of the following work is the development of a user interface for the control system of the heavy ion storage ring TSR at the Max-Planck-Institut for nuclear physics in Heidelberg. The new software has been developed with the programming language Visual Basic TM under the operating system Windows 95 TM . Its purpose is on one hand to enable the operator(s) to calculate some of the parameters for the adjustment of the storage ring on the other hand it improves the means of data acquisition from current, Schottky and BTF (beam transfer function) measurements. To achieve a simple operation of the program has been one of the primary goals during the development of the program. The new user interface was practically tested within three beam times. The collected data was examined to obtain information about momentum spread, longitudinal coupling impedance and minimal detection limit for the number of ions. A measurement with a 32 S 16+ beam rsulted in a limit of about 300. (orig.) [de

Design requirements of communication architecture of SMART safety system

International Nuclear Information System (INIS)

Park, H. Y.; Kim, D. H.; Sin, Y. C.; Lee, J. Y.

2001-01-01

To develop the communication network architecture of safety system of SMART, the evaluation elements for reliability and performance factors are extracted from commercial networks and classified the required-level by importance. A predictable determinacy, status and fixed based architecture, separation and isolation from other systems, high reliability, verification and validation are introduced as the essential requirements of safety system communication network. Based on the suggested requirements, optical cable, star topology, synchronous transmission, point-to-point physical link, connection-oriented logical link, MAC (medium access control) with fixed allocation are selected as the design elements. The proposed architecture will be applied as basic communication network architecture of SMART safety system

Quality assurance requirements in the testing of packages to be used for safe transportation of RAM

International Nuclear Information System (INIS)

Vieru, Gheorghe; Nistor, Viorica; Mihaiu, Ramona

2010-01-01

The quality of the Type A, B or C packages used for transport and storage of Radioactive Material (RAM) has to be proved by performing qualification tests in accordance with the Transport Regulations, within the Reliability and Testing Laboratory, Institute for Nuclear Research (INR) Pitesti, where has designed and developed a new Romanian Testing Facility. The qualifications testing are performed under a strict quality assurance programme based on the specific procedures prior approved by the Romanian Nuclear Regulatory Body CNCAN (National Commission for Nuclear Activity Control). This paper describe the quality assurance programme in accordance with the quality management system developed in order to meet the requirements provided by the national regulations as well as to the requirements of the IAEA's safety standard TS-R-1 related to testing of packages to be used for transport of RAM and also provides an overview of the new Romanian Testing Facilities for RAM Packages, developed by the INR's Reliability and Testing Laboratory within an Excellence Scientific Contract. (authors)

TWRS safety SSCs: Requirements and characteristics

International Nuclear Information System (INIS)

Smith-Fewell, M.A.

1997-01-01

Safety Systems, Structures, and Components (SSCs) have been identified from hazard and accident analyses. These analyses were performed to support the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR) and Basis for Interim Operation (BID). The text identifies and evaluates the SSCs and their supporting SSCs to show that they either prevent the occurrence of the accident or mitigate the consequences of the accident to below the acceptance guidelines. The requirements for the SSCs to fulfill these tasks are described

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-09-15

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

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.

Operating safety requirements for the intermediate level liquid waste system

International Nuclear Information System (INIS)

1980-07-01

The operation of the Intermediate Level Liquid Waste (ILW) System, which is described in the Final Safety Analysis, consists of two types of operations, namely: (1) the operation of a tank farm which involves the storage and transportation through pipelines of various radioactive liquids; and (2) concentration of the radioactive liquids by evaporation including rejection of the decontaminated condensate to the Waste Treatment Plant and retention of the concentrate. The following safety requirements in regard to these operations are presented: safety limits and limiting control settings; limiting conditions for operation; and surveillance requirements. Staffing requirements, reporting requirements, and steps to be taken in the event of an abnormal occurrence are also described

Philosophy and safety requirements for land-based nuclear installations

International Nuclear Information System (INIS)

Kellermann, Otto

1978-01-01

The main ideas of safety philosophy for land-based nuclear installations are presented together with their background of protection goals. Today's requirements for design and quality assurance are deductively shown. Finally a proposition is made for a new balancing of safety philosophy according to the high safety level that nuclear installations have reached

Safety integrity requirements for computer based I ampersand C systems

International Nuclear Information System (INIS)

Thuy, N.N.Q.; Ficheux-Vapne, F.

1997-01-01

In order to take into account increasingly demanding functional requirements, many instrumentation and control (I ampersand C) systems in nuclear power plants are implemented with computers. In order to ensure the required safety integrity of such equipment, i.e., to ensure that they satisfactorily perform the required safety functions under all stated conditions and within stated periods of time, requirements applicable to these equipment and to their life cycle need to be expressed and followed. On the other hand, the experience of the last years has led EDF (Electricite de France) and its partners to consider three classes of systems and equipment, according to their importance to safety. In the EPR project (European Pressurized water Reactor), these classes are labeled E1A, E1B and E2. The objective of this paper is to present the outline of the work currently done in the framework of the ETC-I (EPR Technical Code for I ampersand C) regarding safety integrity requirements applicable to each of the three classes. 4 refs., 2 figs

A comparison of the difference of requirements between functional safety and nuclear safety controllers

Energy Technology Data Exchange (ETDEWEB)

Chen, C.K.; Lee, C.L.; Shyu, S.S. [Inst. of Nuclear Energy Research, Taoyuan, Taiwan (China)

2014-07-01

In order to establish self-reliant capabilities of nuclear I&C systems in Taiwan, Taiwan's Nuclear I&C System (TNICS) project had been established by Institute of Nuclear Energy Research (INER). A Triple Modular Redundant (TMR) safety controller (SCS-2000) has been completed and gone through the IEC 61508 Safety Integrity Level 3 (SIL3) certification of Functional Safety for industries. Based on the certification processes, the difference of requirements between Functional Safety and Nuclear Safety controllers in term of hardware and software are addressed in this study. Besides, the measures used to determine and verify the reliability of the safety control system design are presented. (author)

Predisposal management of radioactive waste. General safety requirements. Pt. 5

International Nuclear Information System (INIS)

2009-01-01

The objective of this Safety Requirements publication is to establish, the requirements that must be satisfied in the predisposal management of radioactive waste. This publication sets out the objectives, criteria and requirements for the protection of human health and the environment that apply to the siting, design, construction, commissioning, operation and shutdown of facilities for the predisposal management of radioactive waste, and the requirements that must be met to ensure the safety of such facilities and activities. This Safety Requirements publication applies to the predisposal management of radioactive waste of all types and covers all the steps in its management from its generation up to its disposal, including its processing (pretreatment, treatment and conditioning), storage and transport. Such waste may arise from the commissioning, operation and decommissioning of nuclear facilities; the use of radionuclides in medicine, industry, agriculture, research and education; the processing of materials that contain naturally occurring radionuclides; and the remediation of contaminated areas. The introduction of the document (Section 1) informs about its objective, scope and structure. The protection of human health and the environment is considered in Section 2 of this publication. Section 3 establishes requirements for the responsibilities associated with the predisposal management of radioactive waste. Requirements for the principal approaches to and the elements of the predisposal management of radioactive waste are established in Section 4. Section 5 establishes requirements for the safe development and operation of predisposal radioactive waste management facilities and safe conduct of activities. The Annex presents a discussion of the consistency of the safety requirements established in this publication with the fundamental safety principles

Recommended general safety requirements for nuclear power plants

International Nuclear Information System (INIS)

1983-06-01

This report presents recommendations for a set of general safety requirements that could form the basis for the licensing of nuclear power plants by the Atomic Energy Control Board. In addition to a number of recommended deterministic requirements the report includes criteria for the acceptability of the design of such plants based upon the calculated probability and consequence (in terms of predicted radiation dose to members of the public) of potential fault sequences. The report also contains a historical review of nuclear safety principles and practices in Canada

Evaluation of safety, an unavoidable requirement in the applications of ionizing radiations

International Nuclear Information System (INIS)

Jova Sed, Luis Andres

2013-01-01

The safety assessments should be conducted as a means to evaluate compliance with safety requirements (and thus the application of fundamental safety principles) for all facilities and activities in order to determine the measures to be taken to ensure safety. It is an essential tool in decision making. For long time we have linked the safety assessment to nuclear facilities and not to all practices involving the use of ionizing radiation in daily life. However, the main purpose of the safety assessment is to determine if it has reached an appropriate level of safety for an installation or activity and if it has fulfilled the objectives of safety and basic safety criteria set by the designer, operating organization and the regulatory body under the protection and safety requirements set out in the International Basic safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. This paper presents some criteria and personal experiences with the new international recommendations on this subject and its practical application in the region and demonstrates the importance of this requirement. Reflects the need to train personnel of the operator and the regulatory body in the proportional application of this requirement in practice with ionizing radiation

Safety of Nuclear Fuel Cycle Facilities. Safety Requirements (Arabic Edition)

International Nuclear Information System (INIS)

2015-01-01

This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific relevance include aspects of nuclear fuel generation, storage, reprocessing and disposal

Risk and safety requirements for diagnostic and therapeutic procedures in allergology

DEFF Research Database (Denmark)

Kowalski, Marek L; Ansotegui, Ignacio; Aberer, Werner

2016-01-01

One of the major concerns in the practice of allergy is related to the safety of procedures for the diagnosis and treatment of allergic disease. Management (diagnosis and treatment) of hypersensitivity disorders involves often intentional exposure to potentially allergenic substances (during skin...... attempted to present general requirements necessary to assure the safety of these procedures. Following review of available literature a group of allergy experts within the World Allergy Organization (WAO), representing various continents and areas of allergy expertise, presents this report on risk...... associated with diagnostic and therapeutic procedures in allergology and proposes a consensus on safety requirements for performing procedures in allergy offices. Optimal safety measures including appropriate location, type and required time of supervision, availability of safety equipment, access...

Requirements to be met by a safety philosophy

International Nuclear Information System (INIS)

Hahn, L.

1990-01-01

The author's assessment of the use of safety philosophies is that, since 'safety philosophers' still are not certain whether a safety philosophy ought to be applicable to just one, particular technology, or rather to a variety of different technologies, there is reason to state that the required ethical, philosophical and political foundations to build a safety philosophy on are still missing. And this, the author presumes, is one of the reasons why our society to a far extent is incapable of acting, faced not only with the nuclear issue, but also with the present and future ecological challenge. (orig./DG) [de

Defence-in-depth and development of safety requirements for advanced nuclear reactors

International Nuclear Information System (INIS)

Carnino, A.; Gasparini, M.

2002-01-01

The paper addresses a general approach for the preparation of the design safety requirements using the IAEA Safety Objectives and the strategy of defence-in-depth. It proposes a general method (top-down approach) to prepare safety requirements for a given kind of reactor using the IAEA requirements for nuclear power plants as a starting point through a critical interpretation and application of the strategy of defence-in-depth. The IAEA has recently developed a general methodology for screening the defence-in-depth of nuclear power plants starting from the fundamental safety objectives as proposed in the IAEA Safety Fundamentals. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor. Currently the IAEA is preparing the technical basis for the development of safety requirements for Modular High Temperature Gas Reactors, with the aim of showing the viability of the method. A draft TECDOC has been prepared and circulated among several experts for comments. This paper is largely based on the content of the draft TECDOC. (authors)

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1, Revision 1 (Chinese Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication establishes requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

EPR meets the next generation PWR safety requirements

International Nuclear Information System (INIS)

Bouteille, Francois; Czech, Juergen; Sloan, Sandra

2006-01-01

At the origin was the common decision in 1989 of Framatome and Siemens to cooperate to design a Nuclear Island which meets the future needs of utilities. EDF and a group of main German Utilities joined this effort in 1991 and from that point were completely involved in the progress of the work. Compliance of the EPR with the European Utility Requirements (EUR) was verified to ensure a large acceptability of the design by other participating utilities. In addition, the entire process was backed up to the end of 1998 by the French and the German Safety Authorities which engaged into a long-lasting cooperation to define common requirements applicable to future Nuclear Power Plants. Upon signature of the Olkiluoto 3 contract, STUK, the Finnish safety and radiation authority, began reviewing the design of the EPR. Upon the favorable recommendation of STUK, the Finnish government delivered a Construction License for the Olkiluoto 3 NPP on February 17, 2005. Following the positive conclusion of the political debate in France with regard to nuclear energy, EDF will also submit a request to start the construction of an EPR on the Flamanville site. In the US, the first steps in view of a Design Certification by the NRC have been taken. These three independent decisions make the EPR the leading first generation 3+ design under construction. Important safety functions are assured by separate systems in a straightforward operating mode. Four separate, redundant trains for all safety systems are installed in four separate layout division for which a strict separation is ensured so that common mode failure, for example due to internal hazards, can be ruled out. A reduction in common mode failure potential is also obtained by design rules ensuring the systematic application of functional diversity. A four train-redundancy for the major safety systems provides flexibility in adapting the design to maintenance requirements, thus contributing to reduce the outage duration. Additional

Predisposal Management of Radioactive Waste. General Safety Requirements Pt. 5

International Nuclear Information System (INIS)

2010-01-01

There are a large number of facilities and activities around the world in which radioactive material is produced, handled and stored. This Safety Requirements publication presents international consensus requirements for the management of radioactive waste prior to its disposal. It provides the safety imperatives on the basis of which facilities can be designed, operated and regulated. The publication is supported by a number of Safety Guides that provide up to date recommendations and guidance on best practices for management of particular types of radioactive waste, for storage of radioactive waste, for assuring safety by developing safety cases and supporting safety assessments, and for applying appropriate management systems. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Responsibilities associated with the predisposal management of radioactive waste; 4. Steps in the predisposal management of radioactive waste; 5. Development and operation of predisposal radioactive waste management facilities and activities; Annex: Predisposal management of radioactive waste and the fundamental safety principles.

Predisposal Management of Radioactive Waste. General Safety Requirements Pt. 5

International Nuclear Information System (INIS)

2009-01-01

There are a large number of facilities and activities around the world in which radioactive material is produced, handled and stored. This Safety Requirements publication presents international consensus requirements for the management of radioactive waste prior to its disposal. It provides the safety imperatives on the basis of which facilities can be designed, operated and regulated. The publication is supported by a number of Safety Guides that provide up to date recommendations and guidance on best practices for management of particular types of radioactive waste, for storage of radioactive waste, for assuring safety by developing safety cases and supporting safety assessments, and for applying appropriate management systems. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Responsibilities associated with the predisposal management of radioactive waste; 4. Steps in the predisposal management of radioactive waste; 5. Development and operation of predisposal radioactive waste management facilities and activities; Annex: Predisposal management of radioactive waste and the fundamental safety principles.

Tank Farms Technical Safety Requirements. Volume 1 and 2

International Nuclear Information System (INIS)

CASH, R.J.

2000-01-01

The Technical Safety Requirements (TSRs) define the acceptable conditions, safe boundaries, basis thereof, and controls to ensure safe operation during authorized activities, for facilities within the scope of the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR)

Tank Farms Technical Safety Requirements [VOL 1 and 2

Energy Technology Data Exchange (ETDEWEB)

CASH, R.J.

2000-12-28

The Technical Safety Requirements (TSRs) define the acceptable conditions, safe boundaries, basis thereof, and controls to ensure safe operation during authorized activities, for facilities within the scope of the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR).

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-04-15

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

Hazard Analysis and Safety Requirements for Small Drone Operations: To What Extent Do Popular Drones Embed Safety?

Science.gov (United States)

Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimihailidou, Maria Mikela

2018-03-01

Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this article presents: (1) a set of safety requirements generated from the application of the systems theoretic process analysis (STPA) technique on a generic small drone system; (2) a gap analysis between the set of safety requirements and the ones met by 19 popular drone models; (3) the extent of the differences between those models, their manufacturers, and the countries of origin; and (4) the association of drone prices with the extent they meet the requirements derived by STPA. The application of STPA resulted in 70 safety requirements distributed across the authority, manufacturer, end user, or drone automation levels. A gap analysis showed high dissimilarities regarding the extent to which the 19 drones meet the same safety requirements. Statistical results suggested a positive correlation between drone prices and the extent that the 19 drones studied herein met the safety requirements generated by STPA, and significant differences were identified among the manufacturers. This work complements the existing risk assessment frameworks for small drones, and contributes to the establishment of a commonly endorsed international risk analysis framework. Such a framework will support the development of a holistic and methodologically justified standardization scheme for small drone flights. © 2017 Society for Risk Analysis.

Safety assessment requirements for onsite transfers of radioactive material

International Nuclear Information System (INIS)

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

1992-05-01

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

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

International Nuclear Information System (INIS)

1992-12-01

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

Safety and environmental requirements and design targets for TIBER-II

International Nuclear Information System (INIS)

Piet, S.J.

1987-09-01

A consistent set of safety and environmental requirements and design targets was proposed and adopted for the TIBER-II (Tokamak Ignition/Burn Experimental Reactor) design effort. TIBER-II is the most recent US version of a fusion experimental test reactor (ETR). These safety and environmental design targets were one contribution of the Fusion Safety Program in the TIBER-II design effort. The other contribution, safety analyses, is documented in the TIBER-II design report. The TIBER-II approach, described here, concentrated on logical development of, first, a complete and consistent set of safety and environmental requirements that are likely appropriate for an ETR, and, second, an initial set of design targets to guide TIBER-II. Because of limited time in the TIBER-II design effort, the iterative process only included one iteration - one set of targets and one design. Future ETR design efforts should therefore build on these design targets and the associated safety analyses. 29 refs., 5 figs., 3 tabs

Canister Storage Building (CSB) Technical Safety Requirements

International Nuclear Information System (INIS)

KRAHN, D.E.

2000-01-01

The purpose of this section is to explain the meaning of logical connectors with specific examples. Logical connectors are used in Technical Safety Requirements (TSRs) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TSRs are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings

A new approach to determine the environmental qualification requirements for the safety related equipment

International Nuclear Information System (INIS)

Hasnaoui, C.; Parent, G.

2000-01-01

The objective of the environmental qualification of safety related equipment is to ensure that the plant defense-in-depth is not compromised by common mode failures following design basis accidents with a harsh environment. A new approach based on safety functions has been developed to determine what safety-related equipment is required to function during and after a design basis accident, as well as their environmental qualification requirements. The main feature of this approach is to use auxiliary safety functions established from safety requirements as credited in the safety analyses. This approach is undertaken in three steps: identification of the auxiliary safety functions of each main safety function; determination of the main equipment groups required for each auxiliary safety function; and review of the safety analyses for design basis accidents in order to determine the credited auxiliary safety functions and their mission times for each accident scenario. Some of the benefits of the proposed approach for the determination of the safety environmental qualification requirements are: a systematic approach for the review of safety analyses based on a safety function check list, and the insurance, with the availability of the safety functions, that Gentilly-2 defense-in-depth would not be compromised by design basis accidents with a harsh environment. (author)

41 CFR 128-1.8006 - Seismic Safety Program requirements.

Science.gov (United States)

2010-07-01

... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Seismic Safety Program requirements. 128-1.8006 Section 128-1.8006 Public Contracts and Property Management Federal Property Management Regulations System (Continued) DEPARTMENT OF JUSTICE 1-INTRODUCTION 1.80-Seismic Safety Program...

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

International Nuclear Information System (INIS)

1993-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. This third annual NUREG report, Supplement 3, presents updated information as of September 30, 1993. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

47 CFR 80.305 - Watch requirements of the Communications Act and the Safety Convention.

Science.gov (United States)

2010-10-01

... and the Safety Convention. 80.305 Section 80.305 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Safety Watch Requirements and Procedures Ship Station Safety Watches § 80.305 Watch requirements of the Communications Act and the Safety...

Safety requirements for the Pu carriers

International Nuclear Information System (INIS)

Mishima, H.

1993-01-01

Ministry of Transport of Japan has now set about studying requirements for Pu carriers to ensure safety. It was first studied what the basic concept of safe carriage of Pu should be, and the basic ideas have been worked out. Next the requirements for the Pu carriers were studied based on the above. There are at present no international requirements of construction and equipment for the nuclear-material carriers, but MOT of Japan has so far required special construction and equipment for the nuclear-material carriers which carry a large amount of radioactive material, such as spent fuel or low level radioactive waste, corresponding to the level of the respective potential hazard. The requirements of construction and equipment of the Pu carriers have been established considering the difference in heat generation between Pu and spent fuel, physical protection, and so forth, in addition to the above basic concept. (J.P.N.)

Quality assurance requirements for the computer software and safety analyses

International Nuclear Information System (INIS)

Husarecek, J.

1992-01-01

The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref

Safety requirements expected to the prototype fast breeder reactor 'Monju'

International Nuclear Information System (INIS)

2014-11-01

In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

Meeting up-to-date safety requirements in the Russian NPP projects

International Nuclear Information System (INIS)

Tepkyan, G. O.; Yashkin, A. V.

2014-01-01

Safety features in Russian NPP designs are implemented by the combination of active and passive safety systems • Russian NPP designs are in compliance with up-to-date international and European safety requirements and refer to Generation III+ • Russian state-of-the-art designs have already implemented some design solutions, which take into account “post-Fukushima” requirements. Russian NPP design principles have been approved during the European discussions in spring 2012, including the IAEA extraordinary session addressed to Fukushima NPP accident

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Spanish Edition)

International Nuclear Information System (INIS)

2010-01-01

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (French Edition)

International Nuclear Information System (INIS)

2010-01-01

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Chinese Edition)

International Nuclear Information System (INIS)

2010-01-01

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Arabic Edition)

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-09-15

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

Specification of advanced safety modeling requirements (Rev. 0).

Energy Technology Data Exchange (ETDEWEB)

Fanning, T. H.; Tautges, T. J.

2008-06-30

The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy established design limits and safety margins under a wide variety of operational, design basis, and beyond design basis transient conditions. Current modeling capabilities for fast reactor safety analyses have resulted from several hundred person-years of code development effort supported by experimental validation. The broad spectrum of mechanistic and phenomenological models that have been developed represent an enormous amount of institutional knowledge that needs to be maintained. Complicating this, the existing code architectures for safety modeling evolved from programming practices of the 1970s. This has lead to monolithic applications with interdependent data models which require significant knowledge of the complexities of the entire code in order for each component to be maintained. In order to develop an advanced fast reactor safety modeling capability, the limitations of the existing code architecture must be overcome while preserving the capabilities that already exist. To accomplish this, a set of advanced safety modeling requirements is defined, based on modern programming practices, that focuses on modular development within a flexible coupling framework. An approach for integrating the existing capabilities of the SAS4A/SASSYS-1 fast reactor safety analysis code into the SHARP framework is provided in order to preserve existing capabilities while providing a smooth transition to advanced modeling capabilities. In doing this, the advanced fast reactor safety models

Specification of advanced safety modeling requirements (Rev. 0)

International Nuclear Information System (INIS)

Fanning, T. H.; Tautges, T. J.

2008-01-01

The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy established design limits and safety margins under a wide variety of operational, design basis, and beyond design basis transient conditions. Current modeling capabilities for fast reactor safety analyses have resulted from several hundred person-years of code development effort supported by experimental validation. The broad spectrum of mechanistic and phenomenological models that have been developed represent an enormous amount of institutional knowledge that needs to be maintained. Complicating this, the existing code architectures for safety modeling evolved from programming practices of the 1970s. This has lead to monolithic applications with interdependent data models which require significant knowledge of the complexities of the entire code in order for each component to be maintained. In order to develop an advanced fast reactor safety modeling capability, the limitations of the existing code architecture must be overcome while preserving the capabilities that already exist. To accomplish this, a set of advanced safety modeling requirements is defined, based on modern programming practices, that focuses on modular development within a flexible coupling framework. An approach for integrating the existing capabilities of the SAS4A/SASSYS-1 fast reactor safety analysis code into the SHARP framework is provided in order to preserve existing capabilities while providing a smooth transition to advanced modeling capabilities. In doing this, the advanced fast reactor safety models will

Current trends in codal requirements for safety in operation of nuclear power plants

International Nuclear Information System (INIS)

Srivasista, K.; Shah, Y.K.; Gupta, S.K.

2006-01-01

The Code of practice on safety in nuclear power plant operation states the requirements to be met during operation of a nuclear power plant for assuring safety. Among various stages of authorization, regulatory body issues authorization for operation of a nuclear power plant, monitors and enforces regulatory requirements. The responsible organization shall have overall responsibility and the plant management shall have the primary responsibility for ensuring safe and efficient operation of its nuclear power plants. A set of codal requirements covering technical and administrative aspects are mandatory for the plant management to implement to ensure that the nuclear power plant is operated in accordance with the design intent. Requirements on operating procedures and instructions establish operation and maintenance, inspection and testing of the plant in a planned and systematic way. The requirements on emergency preparedness programme establish with a reasonable assurance that, in the event of an emergency situation, appropriate measures can be taken to mitigate the consequences. Commissioning requirements verify performance criteria during commissioning to ensure that the design intent and QA requirements are met. Several modifications in systems important to safety required during operation of a nuclear power plant are regulated. However new operational codal requirements arising out of periodic safety review, operational experience feedback, life management, probabilistic safety assessment, physical security, safety convention and obligations and decommissioning are not covered in the present code of practice for safety in nuclear power plant operation. Codal provisions on 'Review by operating organization on aspects of design having implications on operability' are also required to be addressed. The merits in developing such a methodology include acceptance of the design by operating organization, ensuring maintainability, proper layout etc. in the new designs

Safety requirements and safety experience of nuclear facilities in the Federal Republic of Germany

International Nuclear Information System (INIS)

Schnurer, H.L.

1977-01-01

Peaceful use of nuclear energy within the F.R.G. is rapidly growing. The Energy Programme of the Federal Government forecasts a capacity of up to 50.000 MW in 1985. Whereas most of this capacity will be of the LWR-Type, other activities are related to LMFBR - and HTGR - development, nuclear ships, and facilities of the nuclear fuel cycle. Safety of nuclear energy is the pacemaker for the realization of nuclear programmes and projects. Due to a very high population - and industrialisation density, safety has the priority before economical aspects. Safety requirements are therefore extremely stringent, which will be shown for the legal, the technical as well as for the organizational area. They apply for each nuclear facility, its site and the nuclear energy system as a whole. Regulatory procedures differ from many other countries, assigning executive power to state authorities, which are supervised by the Federal Government. Another particularity of the regulatory process is the large scope of involvement of independent experts within the licensing procedures. The developement of national safety requirements in different countries generates a necessity to collaborate and harmonize safety and radiation protection measures, at least for facilities in border areas, to adopt international standards and to assist nuclear developing countries. However, different nationally, regional or local situations might raise problems. Safety experience with nuclear facilities can be concluded from the positive construction and operation experience, including also a few accidents and incidents and the conclusions, which have been drawn for the respective factilities and others of similar design. Another tool for safety assessments will be risk analyses, which are under development by German experts. Final, a scope of future problems and developments shows, that safety of nuclear installations - which has reached a high performance - nevertheless imposes further tasks to be solved

Safety related requirements on future nuclear power plants

International Nuclear Information System (INIS)

Niehaus, F.

1991-01-01

Nuclear power has the potential to significantly contribute to the future energy supply. However, this requires continuous improvements in nuclear safety. Technological advancements and implementation of safety culture will achieve a safety level for future reactors of the present generation of a probability of core-melt of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors there are inherent limitations to increase safety further due to the human element, complexity of design and capability of the containment function. Therefore, revolutionary designs are being explored with the aim of eliminating the potential for off-site releases. In this context it seems to be advisable to explore concepts where the ultimate safety barrier is the fuel itself. (orig.) [de

Development of NPP Safety Requirements into Kenya's Grid Codes

Energy Technology Data Exchange (ETDEWEB)

Ndirangu, Nguni James; Koo, Chang Choong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand.

Development of NPP Safety Requirements into Kenya's Grid Codes

International Nuclear Information System (INIS)

Ndirangu, Nguni James; Koo, Chang Choong

2015-01-01

As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand

DARHT: INTEGRATION OF AUTHORIZATION BASIS REQUIREMENTS AND WORKER SAFETY

International Nuclear Information System (INIS)

MC CLURE, D. A.; NELSON, C. A.; BOUDRIE, R. L.

2001-01-01

This document describes the results of consensus agreements reached by the DARHT Safety Planning Team during the development of the update of the DARHT Safety Analysis Document (SAD). The SAD is one of the Authorization Basis (AB) Documents required by the Department prior to granting approval to operate the DARHT Facility. The DARHT Safety Planning Team is lead by Mr. Joel A. Baca of the Department of Energy Albuquerque Operations Office (DOE/AL). Team membership is drawn from the Department of Energy Albuquerque Operations Office, the Department of Energy Los Alamos Area Office (DOE/LAAO), and several divisions of the Los Alamos National Laboratory. Revision 1 of the DARHT SAD had been written as part of the process for gaining approval to operate the Phase 1 (First Axis) Accelerator. Early in the planning stage for the required update of the SAD for the approval to operate both Phase 1 and Phase 2 (First Axis and Second Axis) DARHT Accelerator, it was discovered that a conflict existed between the Laboratory approach to describing the management of facility and worker safety

Nuclear fuels with high burnup: safety requirements

International Nuclear Information System (INIS)

Phuc Tran Dai

2016-01-01

Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

Fuel Supply Shutdown Facility Interim Operational Safety Requirements

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

Preparedness and response for a nuclear or radiological emergency. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

This Safety Requirements publication establishes the requirements for an adequate level of preparedness and response for a nuclear or radiological emergency in any State. Their implementation is intended to minimize the consequences for people, property and the environment of any nuclear or radiological emergency. The fulfilment of these requirements will also contribute to the harmonization of arrangements in the event of a transnational emergency. These requirements are intended to be applied by authorities at the national level by means of adopting legislation, establishing regulations and assigning responsibilities. The requirements apply to all those practices and sources that have the potential for causing radiation exposure or environmental radioactive contamination warranting an emergency intervention and that are: (a) Used in a State that chooses to adopt the requirements or that requests any of the sponsoring organizations to provide for the application of the requirements. (B) Used by States with the assistance of the FAO, IAEA, ILO, PAHO, OCHA or WHO in compliance with applicable national rules and regulations. (C) Used by the IAEA or which involve the use of materials, services, equipment, facilities and non-published information made available by the IAEA or at its request or under its control or supervision. Or (d) Used under any bilateral or multilateral arrangement whereby the parties request the IAEA to provide for the application of the requirements. The requirements also apply to the off-site jurisdictions that may need to make an emergency intervention in a State that adopts the requirements. The types of practices and sources covered by these requirements include: fixed and mobile nuclear reactors. Facilities for the mining and processing of radioactive ores. Facilities for fuel reprocessing and other fuel cycle facilities. Facilities for the management of radioactive waste. The transport of radioactive material. Sources of radiation used in

42 CFR 3.210 - Required disclosure of patient safety work product to the Secretary.

Science.gov (United States)

2010-10-01

... 42 Public Health 1 2010-10-01 2010-10-01 false Required disclosure of patient safety work product... HUMAN SERVICES GENERAL PROVISIONS PATIENT SAFETY ORGANIZATIONS AND PATIENT SAFETY WORK PRODUCT Confidentiality and Privilege Protections of Patient Safety Work Product § 3.210 Required disclosure of patient...

Discussion of important safety requirements for new nuclear power plants

International Nuclear Information System (INIS)

Zhang Lin; Jia Xiang; Yan Tianwen; Li Wenhong; Li Chun

2014-01-01

This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement. (authors)

Development of photovoltaic array and module safety requirements

Science.gov (United States)

1982-01-01

Safety requirements for photovoltaic module and panel designs and configurations likely to be used in residential, intermediate, and large-scale applications were identified and developed. The National Electrical Code and Building Codes were reviewed with respect to present provisions which may be considered to affect the design of photovoltaic modules. Limited testing, primarily in the roof fire resistance field was conducted. Additional studies and further investigations led to the development of a proposed standard for safety for flat-plate photovoltaic modules and panels. Additional work covered the initial investigation of conceptual approaches and temporary deployment, for concept verification purposes, of a differential dc ground-fault detection circuit suitable as a part of a photovoltaic array safety system.

The Management System for Facilities and Activities. Safety Requirements

International Nuclear Information System (INIS)

2011-01-01

This publication establishes requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States in establishing and implementing effective management systems that integrate all aspects of managing nuclear facilities and activities in a coherent manner. It details the planned and systematic actions necessary to provide adequate confidence that all these requirements are satisfied. Contents: 1. Introduction; 2. Management system; 3. Management responsibility; 4. Resource management; 5. Process implementation; 6. Measurement, assessment and improvement.

Safety Design Requirements for The Interior Architecture of Scientific Research Laboratories

International Nuclear Information System (INIS)

ElDib, A.A.

2014-01-01

The paper discusses one of the primary objectives of interior architecture design of research laboratories (specially those using radioactive materials) where it should provide a safe, accessible environment for laboratory personnel to conduct their work. A secondary objective is to allow for maximum flexibility for safe research. Therefore, health and safety hazards must be anticipated and carefully evaluated so that protective measures can be incorporated into the interior architectural design of these facilities wherever possible. The interior architecture requirements discussed in this paper illustrate some of the basic health and safety design features required for new and remodeled laboratories.The paper discusses one of the primary objectives of interior architecture design of research laboratories (specially those using radioactive materials) where it should provide a safe, accessible environment for laboratory personnel to conduct their work. A secondary objective is to allow for maximum flexibility for safe research. Therefore, health and safety hazards must be anticipated and carefully evaluated so that protective measures can be incorporated into the interior architectural design of these facilities wherever possible. The interior architecture requirements discussed in this paper illustrate some of the basic health and safety design features required for new and remodeled laboratories.

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-03-03

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

Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues. Supplement 4

International Nuclear Information System (INIS)

1994-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. Supplement 3 gives status as of September 30, 1993. This annual report, Supplement 4, presents updated information as of September 30, 1994. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

Software Safety Analysis of Digital Protection System Requirements Using a Qualitative Formal Method

International Nuclear Information System (INIS)

Lee, Jang-Soo; Kwon, Kee-Choon; Cha, Sung-Deok

2004-01-01

The safety analysis of requirements is a key problem area in the development of software for the digital protection systems of a nuclear power plant. When specifying requirements for software of the digital protection systems and conducting safety analysis, engineers find that requirements are often known only in qualitative terms and that existing fault-tree analysis techniques provide little guidance on formulating and evaluating potential failure modes. A framework for the requirements engineering process is proposed that consists of a qualitative method for requirements specification, called the qualitative formal method (QFM), and a safety analysis method for the requirements based on causality information, called the causal requirements safety analysis (CRSA). CRSA is a technique that qualitatively evaluates causal relationships between software faults and physical hazards. This technique, extending the qualitative formal method process and utilizing information captured in the state trajectory, provides specific guidelines on how to identify failure modes and the relationship among them. The QFM and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital protection system example

The actual development of European aviation safety requirements in aviation medicine: prospects of future EASA requirements.

Science.gov (United States)

Siedenburg, J

2009-04-01

Common Rules for Aviation Safety had been developed under the aegis of the Joint Aviation Authorities in the 1990s. In 2002 the Basic Regulation 1592/2002 was the founding document of a new entity, the European Aviation Safety Agency. Areas of activity were Certification and Maintenance of aircraft. On 18 March the new Basic Regulation 216/2008, repealing the original Basic Regulation was published and applicable from 08 April on. The included Essential Requirements extended the competencies of EASA inter alia to Pilot Licensing and Flight Operations. The future aeromedical requirements will be included as Annex II in another Implementing Regulation on Personnel Licensing. The detailed provisions will be published as guidance material. The proposals for these provisions have been published on 05 June 2008 as NPA 2008- 17c. After public consultation, processing of comments and final adoption the new proposals may be applicable form the second half of 2009 on. A transition period of four year will apply. Whereas the provisions are based on Joint Aviation Requirement-Flight Crew Licensing (JAR-FCL) 3, a new Light Aircraft Pilot Licence (LAPL) project and the details of the associated medical certification regarding general practitioners will be something new in aviation medicine. This paper consists of 6 sections. The introduction outlines the idea of international aviation safety. The second section describes the development of the Joint Aviation Authorities (JAA), the first step to common rules for aviation safety in Europe. The third section encompasses a major change as next step: the foundation of the European Aviation Safety Agency (EASA) and the development of its rules. In the following section provides an outline of the new medical requirements. Section five emphasizes the new concept of a Leisure Pilot Licence. The last section gives an outlook on ongoing rulemaking activities and the opportunities of the public to participate in them.

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

International Nuclear Information System (INIS)

Vovk, Ivan

1998-01-01

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

Recommended safety objectives, principles and requirements for mini-reactors

International Nuclear Information System (INIS)

1991-05-01

Canadian and international publications containing objectives, principles and requirements for the safety of nuclear facilities in general and nuclear power plants in particular have been reviewed for their relevance to mini-reactors. Most of the individual recommendations, sometimes with minor wording changes, are applicable to mini-reactors. However, some prescriptive requirements for the shutdown, emergency core cooling and containment systems of power reactors are considered inappropriate for mini-reactors. The Advisory Committee on Nuclear Safety favours a generally non-prescriptive approach whereby the applicant for a mini-reactor license is free to propose any means of satisfying the fundamental objectives, but must convince the regulatory agency to that effect. To do so, a probabilistic safety assessment (PSA) would be the favoured procedure. A generic PSA for all mini-reactors of the same design would be acceptable. Notwithstanding this non-prescriptive approach, the ACNS considers that it would be prudent to require the existence of at least one independent shutdown system and two physically independent locations from which the reactor can be shut down and the shutdown condition monitored, and to require provision for an assumed loss of integrity of the primary cooling system's boundary unless convincing arguments to the contrary are presented. The ACNS endorses in general the objectives and fundamental principles proposed by the interorganizational Small Reactor Criteria working group, and intends to review and comment on the documents on specific applications to be issued by that working group

Safety requirements for long term operation of NPPs

International Nuclear Information System (INIS)

Houdre, T.; Osouf, N.; Juvin, J.-C.

2012-01-01

In the future, the reactors operating at present will run alongside reactors of the EPR type or their equivalent, designed for a significantly higher level of safety. This raises the question of the acceptability of continued operation of reactors beyond 40 years when there is an available technology that is safer. Two objectives are therefore imperative. First, a re-evaluation of the safety level in the light of that required of EPR type reactors or their equivalent is necessary, with proposals to bring about significant and relevant improvements to the reactors. R and D work in France and elsewhere is already indicating orientations that could lead to answers, and improvements that would provide significant reductions in release in case of severe accident are being studied. Second, strict compliance of the reactors with the applicable regulations must be demonstrated. At the same time, ageing and obsolescence of the equipment will have to be managed. Where these two points are concerned, ASN expects far-reaching proposals from the licensee. With a view to a request for continued operation beyond 40 years, ASN has referred the matter to the Advisory Committee for nuclear reactors which will meet at the end of 2011 to establish the safety requirements for reactors at their fourth ten-yearly outage. (author)

Fuel supply shutdown facility interim operational safety requirements

International Nuclear Information System (INIS)

Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

1995-01-01

These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

Determination of the neutron energy and spatial distributions of the neutron beam from the TSR-II in the large beam shield

International Nuclear Information System (INIS)

Clifford, C.E.; Muckenthaler, F.J.

1976-01-01

The TSR-II reactor of the ORNL Tower Shielding Facility has recently been relocated within a new, fixed shield. A principal feature of the new shield is a beam port of considerably larger area than that of its predecessor. The usable neutron flux has thereby been increased by a factor of approximately 200. The bare beam neutron spectrum behind the new shield has been experimentally determined over the energy range from 0.8 to 16 MeV. A high level of fission product gamma ray background prevented measurement of bare beam spectra below 0.8 MeV, however neutron spectra in the energy range from 8 keV to 1.4 MeV were obtained for two simple, calculable shielding configurations. Also measured in the present work were weighted integral flux distributions and fast neutron dose rates

Safety and regulatory requirements of nuclear power plants

International Nuclear Information System (INIS)

Kumar, S.V.; Bhardwaj, S.A.

2000-01-01

A pre-requisite for a nuclear power program in any country is well established national safety and regulatory requirements. These have evolved for nuclear power plants in India with participation of the regulatory body, utility, research and development (R and D) organizations and educational institutions. Prevailing international practices provided a useful base to develop those applicable to specific system designs for nuclear power plants in India. Their effectiveness has been demonstrated in planned activities of building up the nuclear power program as well as with unplanned activities, like those due to safety related incidents etc. (author)

Technical Review Report for the Model 9975-96 Package Safety Analysis Report for Packaging (S-SARP-G-00003, Revision 0, January 2008)

International Nuclear Information System (INIS)

West, M.

2009-01-01

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the U.S. Department of Energy (DOE), on the Safety Analysis Report for Packaging, Model 9975, Revision 0, dated January 2008 (S-SARP-G-00003, the SARP). The review includes an evaluation of the SARP, with respect to the requirements specified in 10 CFR 71, and in International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9975-96 Package is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The safety analysis of the Model 9975-85 Packaging is documented in the Safety Analysis Report for Packaging, Model 9975, B(M)F-85, Revision 0, dated December 2003. The Model 9975-85 Package is certified by DOE Certificate of Compliance (CoC) package identification number, USA/9975/B(M)F-85, for the transportation of Type B quantities of uranium metal/oxide, 238 Pu heat sources, plutonium/uranium metals, plutonium/uranium oxides, plutonium composites, plutonium/tantalum composites, 238 Pu oxide/beryllium metal.

The influence of petrography, mineralogy and chemistry on burnability and reactivity of quicklime produced in Twin Shaft Regenerative (TSR) kilns from Neoarchean limestone (Transvaal Supergroup, South Africa)

Science.gov (United States)

Vola, Gabriele; Sarandrea, Luca; Della Porta, Giovanna; Cavallo, Alessandro; Jadoul, Flavio; Cruciani, Giuseppe

2017-12-01

This study evaluates the influence of chemical, mineralogical and petrographic features of the Neoarchean limestone from the Ouplaas Mine (Griqualand West, South Africa) on its burnability and quicklime reactivity, considering the main use as raw material for high-grade lime production in twin shaft regenerative (TSR) kilns. This limestone consists of laminated clotted peloidal micrite and fenestrate microbial boundstone with herringbone calcite and organic carbon (kerogen) within stylolites. Diagenetic modifications include hypidiotopic dolomite, micrite to microsparite recrystallization, stylolites, poikilotopic calcite, chert and saddle dolomite replacements. Burning and technical tests widely attest that the Neoarchean limestone is sensitive to high temperature, showing an unusual and drastically pronounced sintering or overburning tendency. The slaking reactivity, according to EN 459-2 is high for lime burnt at 1050 °C, but rapidly decreases for lime burnt at 1150 °C. The predominant micritic microbial textures, coupled with the organic carbon, are key-factors influencing the low burnability and the high sintering tendency. The presence of burial cementation, especially poikilotopic calcite, seems to promote higher burnability, either in terms of starting calcination temperature, or in terms of higher carbonate dissociation rate. In fact, the highest calcination velocity determined by thermal analysis is consistent with the highest slaking reactivity of the lower stratum of the quarry, enriched in poikilotopic calcite. Secondly, locally concentered dolomitic marly limestones, and sporadic back shales negatively affects the quicklime reactivity, as well. This study confirms that a multidisciplinary analytical approach is essential for selecting the best raw mix for achieving the highest lime reactivity in TSR kilns.

Requirements on the provisional safety analyses and technical comparison of safety measures

International Nuclear Information System (INIS)

2010-04-01

decide on the provision of a design license for a repository site for SMA and another one for HAA, or for a common site for both SMA and HAA. The present report concerns the second step and recapitulates the assertions of SGT on the provisional safety analyses and on the safety technical comparison. It establishes the specific requirements of the Swiss Federal Nuclear Safety Inspectorate (ENSI) on provisional safety and the safety technical comparison. Further, it defines the extent and content of the safety technical documentation necessary for step 2

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

International Nuclear Information System (INIS)

2000-01-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

Workshop on Program for Elimination of Requirements Marginal to Safety: Proceedings

International Nuclear Information System (INIS)

Dey, M.

1993-09-01

These are the proceedings of the Public Workshop on the US Nuclear Regulatory Commission's Program for Elimination of Requirements Marginal to Safety. The workshop was held at the Holiday Inn, Bethesda, on April 27 and 28, 1993. The purpose of the workshop was to provide an opportunity for public and industry input to the program. The workshop addressed the institutionalization of the program to review regulations with the purpose of eliminating those that are marginal. The objective is to avoid the dilution of safety efforts. One session was devoted to discussion of the framework for a performance-based regulatory approach. In addition, panelists and attendees discussed scope, schedules and status of specific regulatory items: containment leakage testing requirements, fire protection requirements, requirements for environmental qualification of electrical equipment, requests for information under 10CFR50.54(f), requirements for combustible gas control systems, and quality assurance requirements

Workshop on Program for Elimination of Requirements Marginal to Safety: Proceedings

Energy Technology Data Exchange (ETDEWEB)

Dey, M. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution; Arsenault, F.; Patterson, M.; Gaal, M. [SCIENTECH, Inc., Rockville, MD (United States)

1993-09-01

These are the proceedings of the Public Workshop on the US Nuclear Regulatory Commission`s Program for Elimination of Requirements Marginal to Safety. The workshop was held at the Holiday Inn, Bethesda, on April 27 and 28, 1993. The purpose of the workshop was to provide an opportunity for public and industry input to the program. The workshop addressed the institutionalization of the program to review regulations with the purpose of eliminating those that are marginal. The objective is to avoid the dilution of safety efforts. One session was devoted to discussion of the framework for a performance-based regulatory approach. In addition, panelists and attendees discussed scope, schedules and status of specific regulatory items: containment leakage testing requirements, fire protection requirements, requirements for environmental qualification of electrical equipment, requests for information under 10CFR50.54(f), requirements for combustible gas control systems, and quality assurance requirements.

Preparation, review, and approval of implementation plans for nuclear safety requirements

International Nuclear Information System (INIS)

1994-10-01

This standard describes an acceptable method to prepare, review, and approve implementation plans for DOE Nuclear Safety requirements. DOE requirements are identified in DOE Rules, Orders, Notices, Immediate Action Directives, and Manuals

Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements. Pt. 3 (Chinese Edition)

International Nuclear Information System (INIS)

2014-01-01

This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

Radiation protection and safety of radiation sources: International basic safety standards. General safety requirements. Pt. 3 (French Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements. Pt. 3 (Arabic Edition)

International Nuclear Information System (INIS)

2015-01-01

This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

Radiation safety requirements for training of users of diagnostic X ...

African Journals Online (AJOL)

Background. Globally, the aim of requirements regarding the use and ownership of diagnostic medical X-ray equipment is to limit radiation by abiding by the 'as low as reasonably achievable' (ALARA) principle. The ignorance of radiographers with regard to radiation safety requirements, however, is currently a cause of ...

Correct safety requirements during the life cycle of heating plants; Korrekta saekerhetskrav under vaermeanlaeggningars livscykel

Energy Technology Data Exchange (ETDEWEB)

Tegehall, Jan; Hedberg, Johan [Swedish National Testing and Research Inst., Boraas (Sweden)

2006-10-15

The safety of old steam boilers or hot water generators is in principle based on electromechanical components which are generally easy to understand. The use of safety-PLC is a new and flexible way to design a safe system. A programmable system offers more degrees of freedom and consequently new problems may arise. As a result, new standards which use the Safety Integrity Level (SIL) concept for the level of safety have been elaborated. The goal is to define a way of working to handle requirements on safety in control systems of heat and power plants. SIL-requirements are relatively new within the domain and there is a need for guidance to be able to follow the requirements. The target of this report is the people who work with safety questions during new construction, reconstruction, or modification of furnace plants. In the work, the Pressure Equipment Directive, 97/23/EC, as well as standards which use the SIL concept have been studied. Additionally, standards for water-tube boilers have been studied. The focus has been on the safety systems (safety functions) which are used in water-tube boilers for heat and power plants; other systems, which are parts of these boilers, have not been considered. Guidance has been given for the aforementioned standards as well as safety requirements specification and risk analysis. An old hot water generator and a relatively new steam boiler have been used as case studies. The design principles and safety functions of the furnaces have been described. During the risk analysis important hazards were identified. A method for performing a risk analysis has been described and the appropriate content of a safety requirements specification has been defined. If a heat or power plant is constructed, modified, or reconstructed, a safety life cycle shall be followed. The purpose of the safety life cycle is to plan, describe, document, perform, check, test, and validate that everything is correctly done. The components of the safety

Results Of Physicochemical Characterization And Caustic Dissolution Tests On Tank 241-C-108 Heel Solids

International Nuclear Information System (INIS)

Callaway, W.S.; Huber, H.J.

2010-01-01

Based on an ENRAF waste surface measurement taken February 1, 2009, double-shell tank (DST) 241-AN-106 (AN-106) contained approximately 278.98 inches (793 kgal) of waste. A zip cord measurement from the tank on February 1, 2009, indicated a settled solids layer of 91.7 inches in height (280 kgal). The supernatant layer in February 2009, by difference, was approximately 187 inches deep (514 kgal). Laboratory results from AN-106 February 1, 2009 (see Table 2) grab samples indicated the supernatant was below the chemistry limit that applied at the time as identified in HNF-SD-WM-TSR-006, Tank Farms Technical Safety Requirements, Administrative Control (AC) 5.16, 'Corrosion Mitigation Controls.' (The limits have since been removed from the Technical Safety Requirements (TSR) and are captured in OSD-T-151-00007, Operating Specifications for the Double-Shell Storage Tanks.) Problem evaluation request WRPS-PER-2009-0218 was submitted February 9, 2009, to document the finding that the supernatant chemistry for grab samples taken from the middle and upper regions of the supernatant was noncompliant with the chemistry control limits. The lab results for the samples taken from the bottom region of the supernatant met AC 5.16 limits.

RESULTS OF PHYSICOCHEMICAL CHARACTERIZATION AND CAUSTIC DISSOLUTION TESTS ON TANK 241-C-108 HEEL SOLIDS

Energy Technology Data Exchange (ETDEWEB)

CALLAWAY WS; HUBER HJ

2010-07-01

Based on an ENRAF waste surface measurement taken February 1, 2009, double-shell tank (DST) 241-AN-106 (AN-106) contained approximately 278.98 inches (793 kgal) of waste. A zip cord measurement from the tank on February 1, 2009, indicated a settled solids layer of 91.7 inches in height (280 kgal). The supernatant layer in February 2009, by difference, was approximately 187 inches deep (514 kgal). Laboratory results from AN-106 February 1, 2009 (see Table 2) grab samples indicated the supernatant was below the chemistry limit that applied at the time as identified in HNF-SD-WM-TSR-006, Tank Farms Technical Safety Requirements, Administrative Control (AC) 5.16, 'Corrosion Mitigation Controls.' (The limits have since been removed from the Technical Safety Requirements (TSR) and are captured in OSD-T-151-00007, Operating Specifications for the Double-Shell Storage Tanks.) Problem evaluation request WRPS-PER-2009-0218 was submitted February 9, 2009, to document the finding that the supernatant chemistry for grab samples taken from the middle and upper regions of the supernatant was noncompliant with the chemistry control limits. The lab results for the samples taken from the bottom region of the supernatant met AC 5.16 limits.

77 FR 75439 - Guidances for Industry and Investigators on Safety Reporting Requirements for Investigational New...

Science.gov (United States)

2012-12-20

...] Guidances for Industry and Investigators on Safety Reporting Requirements for Investigational New Drug Applications and Bioavailability/Bioequivalence Studies, and a Small Entity Compliance Guide; Availability... Reporting Requirements for INDs and BA/BE Studies'' and ``Safety Reporting Requirements for INDs and BA/BE...

76 FR 50331 - Hazardous Materials Regulations; Compatibility With the Regulations of the International Atomic...

Science.gov (United States)

2011-08-12

... geometry requirements applicable to tested fissile material packages. This TS-R-1 change is applicable to... percussion test.) The TS-R-1 revisions pertaining to the solar insolation conditions to be assumed in...

Modeling of requirement specification for safety critical real time computer system using formal mathematical specifications

International Nuclear Information System (INIS)

Sankar, Bindu; Sasidhar Rao, B.; Ilango Sambasivam, S.; Swaminathan, P.

2002-01-01

Full text: Real time computer systems are increasingly used for safety critical supervision and control of nuclear reactors. Typical application areas are supervision of reactor core against coolant flow blockage, supervision of clad hot spot, supervision of undesirable power excursion, power control and control logic for fuel handling systems. The most frequent cause of fault in safety critical real time computer system is traced to fuzziness in requirement specification. To ensure the specified safety, it is necessary to model the requirement specification of safety critical real time computer systems using formal mathematical methods. Modeling eliminates the fuzziness in the requirement specification and also helps to prepare the verification and validation schemes. Test data can be easily designed from the model of the requirement specification. Z and B are the popular languages used for modeling the requirement specification. A typical safety critical real time computer system for supervising the reactor core of prototype fast breeder reactor (PFBR) against flow blockage is taken as case study. Modeling techniques and the actual model are explained in detail. The advantages of modeling for ensuring the safety are summarized

Nuclear safety review requirements for launch approval

International Nuclear Information System (INIS)

Sholtis, J.A. Jr.; Winchester, R.O.

1992-01-01

Use of nuclear power systems in space requires approval which is preceded by extensive safety analysis and review. This careful study allows an informed risk-benefit decision at the highest level of our government. This paper describes the process as it has historically been applied to U.S. isotopic power systems. The Ulysses mission, launched in October 1990, is used to illustrate the process. Expected variations to deal with reactor-power systems are explained

Romania - NPP PLiM Between Regulatory Requirement / Oversight and Operator Safety / Financial Interest

International Nuclear Information System (INIS)

Goicea, Lucian

2012-01-01

Cernavoda Unit 1 PLiM started in the first third of its design life, to develop as regulatory requirements of the components of standards and programmes and to benefit by earlier implementation of the measures for achieving maximum operating life. CNCAN regulatory present approach on the utility PLiM combines the regulatory requirements on management system, ageing management provisions of periodic safety review, detailed technical requirements of ageing programmes and different techniques focusing only on safety issues. (author)

Specification of safety requirements for waste packages with respect to practicable quality control measures

International Nuclear Information System (INIS)

Gruendler, D.; Wurtinger, W.

1987-01-01

Waste packages for disposal in a repository in the Federal Republic of Germany have to meet safety requirements derived from site specific safety analyses. The examination of the waste packages with regard to compliance with these requirements is the main objective of quality control measures. With respect to quality control the requirements have to be specified in a way that practicable control measures can be applied. This is dealt with for the quality control of the activity inventory and the quality control of the waste form. The paper discusses the determination of the activity of hard-to-measure radionuclides and the specification of safety related requirements for the waste form and the packaging using typical examples

Ferrocyanide Safety Program: Data requirements for the ferrocyanide safety issue developed through the data quality objectives (DQO) process

International Nuclear Information System (INIS)

Buck, J.W.; Anderson, C.M.; Pulsipher, B.A.; Toth, J.J.; Turner, P.J.; Cash, R.J.; Dukelow, G.T.; Meacham, J.E.

1993-12-01

This document records the data quality objectives (DQO) process applied to the Ferrocyanide Waste Tank Safety Issue at the Hanford Site by the Pacific Northwest Laboratory and Westinghouse Hanford Company. Specifically, the major recommendations and findings from this Ferrocyanide DQO process are presented so that decision makers can determine the type, quantity, and quality of data required for addressing tank safety issues. The decision logic diagrams and error tolerance equations also are provided. Finally, the document includes the DQO sample-size formulas for determining specific tank sampling requirements

Innovative nuclear reactor - Indian approach to meet user requirements for safety

International Nuclear Information System (INIS)

Saha, D.; Sinha, R.K.

2002-01-01

Full text: For sustainable development of nuclear energy, a number of key issues are to be addressed. It should be economically competitive; it must address the issues related to nuclear safety, proliferation resistance, environmental impact, waste disposal and cross cutting issues like social and infra-structural aspects. To compete successfully in the long term, in the highly competitive energy market and to overcome other challenges, it is necessary to introduce innovative reactor and fuel cycle concepts. Indian Advanced Heavy Water Reactor (AHWR) is one such innovative reactor. To guide the research and development activities related to innovative concepts, user requirements are to be formulated. User requirements covering various aspects of sustainable development are being formulated at both national and international levels. One such international project involved in the formulation of user requirements is the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper deals with INPRO user requirements for safety and Indian approach to meet these requirements through AHWR

42 CFR 9.10 - Occupational Health and Safety Program (OHSP) and biosafety requirements.

Science.gov (United States)

2010-10-01

... 42 Public Health 1 2010-10-01 2010-10-01 false Occupational Health and Safety Program (OHSP) and... SANCTUARY SYSTEM § 9.10 Occupational Health and Safety Program (OHSP) and biosafety requirements. (a) How are employee Occupational Health and Safety Program risks and concerns addressed? The sanctuary shall...

A Review of Safety and Design Requirements of the Artificial Pancreas.

Science.gov (United States)

Blauw, Helga; Keith-Hynes, Patrick; Koops, Robin; DeVries, J Hans

2016-11-01

As clinical studies with artificial pancreas systems for automated blood glucose control in patients with type 1 diabetes move to unsupervised real-life settings, product development will be a focus of companies over the coming years. Directions or requirements regarding safety in the design of an artificial pancreas are, however, lacking. This review aims to provide an overview and discussion of safety and design requirements of the artificial pancreas. We performed a structured literature search based on three search components-type 1 diabetes, artificial pancreas, and safety or design-and extended the discussion with our own experiences in developing artificial pancreas systems. The main hazards of the artificial pancreas are over- and under-dosing of insulin and, in case of a bi-hormonal system, of glucagon or other hormones. For each component of an artificial pancreas and for the complete system we identified safety issues related to these hazards and proposed control measures. Prerequisites that enable the control algorithms to provide safe closed-loop control are accurate and reliable input of glucose values, assured hormone delivery and an efficient user interface. In addition, the system configuration has important implications for safety, as close cooperation and data exchange between the different components is essential.

Information Management system of the safety regulatory requirements and guidance for the Korea next generation reactors

International Nuclear Information System (INIS)

Yun, Y. C.; Lee, J. H.; Lee, H. C.; Lee, J. S.

2000-01-01

In order to achieve the safety of the Korea Next Generation Reactors (KNGR), the Korea Institute of Nuclear Safety has carried out the Safety and Regulatory Requirements and Guidance (SRRG) development program from 1992 such as establishment of the SRRG hierarchy, development of technical requirements and guidance, and consideration of new licensing system. The SRRG hierarchy for the KNGR was consisted of five tiers; Safety Objectives, Safety Principles, General Safety Criteria, Specific Safety Requirements and Safety Regulatory Guides. The developed SRRG have been compared the criteria in 10CFR and Reg. Guide in the U.S.A and the IAEA documents for assuring internationally acceptable level of the SRRG. To improve the efficiency and accuracy of SRRG development, the construction of database system was required in the course of development. Therefore, the Information Management System of SRRG for the KNGR has been developed which enables developers to quickly and accurately seek and systematically manage whole contexts of the SRRG, reference requirements, and current atomic energy regulation rules. Moreover, through homepage whose URL is 'http://kngr.kins.re.kr', the concerned persons and public can acquire the information related with SRRG and KNGR project, and post his/her thought to the opinion forum in the homepage

Information Management system of the safety regulatory requirements and guidance for the Korea next generation reactors

Energy Technology Data Exchange (ETDEWEB)

Yun, Y. C. [LG-EDS Systems, Seoul (Korea, Republic of); Lee, J. H.; Lee, H. C.; Lee, J. S. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

2000-05-01

In order to achieve the safety of the Korea Next Generation Reactors (KNGR), the Korea Institute of Nuclear Safety has carried out the Safety and Regulatory Requirements and Guidance (SRRG) development program from 1992 such as establishment of the SRRG hierarchy, development of technical requirements and guidance, and consideration of new licensing system. The SRRG hierarchy for the KNGR was consisted of five tiers; Safety Objectives, Safety Principles, General Safety Criteria, Specific Safety Requirements and Safety Regulatory Guides. The developed SRRG have been compared the criteria in 10CFR and Reg. Guide in the U.S.A and the IAEA documents for assuring internationally acceptable level of the SRRG. To improve the efficiency and accuracy of SRRG development, the construction of database system was required in the course of development. Therefore, the Information Management System of SRRG for the KNGR has been developed which enables developers to quickly and accurately seek and systematically manage whole contexts of the SRRG, reference requirements, and current atomic energy regulation rules. Moreover, through homepage whose URL is 'http://kngr.kins.re.kr', the concerned persons and public can acquire the information related with SRRG and KNGR project, and post his/her thought to the opinion forum in the homepage.

49 CFR 1106.3 - Actions for which Safety Integration Plan is required.

Science.gov (United States)

2010-10-01

... 49 Transportation 8 2010-10-01 2010-10-01 false Actions for which Safety Integration Plan is required. 1106.3 Section 1106.3 Transportation Other Regulations Relating to Transportation (Continued... TRANSPORTATION BOARD CONSIDERATION OF SAFETY INTEGRATION PLANS IN CASES INVOLVING RAILROAD CONSOLIDATIONS...

Regulatory requirements and administrative practice in safety of nuclear installations

International Nuclear Information System (INIS)

Servant, J.

1977-01-01

This paper reviews the current situation of the France regulatory rules and procedures dealing with the safety of the main nuclear facilities and, more broadly, the nuclear security. First, the author outlines the policy of the French administration which requires that the licensee responsible for an installation has to demonstrate that all possible measures are taken to ensure a sufficient level of safety, from the early stage of the project to the end of the operation of the plant. Thus, the administration performs the assessment on a case-by-case basis, of the safety of each installation before granting a nuclear license. On the other hand, the administration settles overall safety requirements for specific categories of installations or components, which determine the ultimate safety performances, but avoid, as far as possible, to detail the technical specifications to be applied in order to comply with these goals. This approach, which allows the designers and the licensees to rely upon sound codes and standards, gains the advantage of a great flexibility without imparing the nuclear safety. The author outlines the licensing progress for the main categories of installations: nuclear power plants of the PWR type, fast breeders, uranium isotope separation plants, and irradiated fuel processing plants. Emphasis is placed on the most noteworthy points: standardization of projects, specific risks of each site, problems of advanced type reactors, etc... The development of the technical regulations is presented with emphasis on the importance of an internationally concerned action within the nuclear international community. The second part of this paper describes the France operating experience of nuclear installations from the safety point of view. Especially, the author examines the technical and administrative utilization of data from safety significant incidents in reactors and plants, and the results of the control performed by the nuclear installations

Guide for reviewing safety analysis reports for packaging: Review of quality assurance requirements

International Nuclear Information System (INIS)

Moon, D.W.

1988-10-01

This review section describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging which are important to safety. The design effort, operation's plans, and quality assurance requirements should be integrated to achieve a system in which the independent QA program is not overly stringent and the application of QA requirements is commensurate with safety significance. The reviewer must verify that the applicant's QA section in the SARP contains package-specific QA information required by DOE Orders and federal regulations that demonstrate compliance. 8 refs

Impact of New Radiation Safety Standards on Licensing Requirements of Nuclear Power Plant

International Nuclear Information System (INIS)

Strohal, P.; Subasic, D.; Valcic, I.

1996-01-01

As the outcomes of the newly introduced safety philosophies, new and more strict safety design requirements for nuclear installation are expected to be introduced. New in-depth defence measures should be incorporated into the design and operation procedure for a nuclear installation, to compensate for potential failures in protection or safety measures. The new requirements will also apply to licensing of NPP's operation as well as to licensing of nuclear sites, especially for radioactive waste disposal sites. This paper intends to give an overview of possible impacts of new internationally agreed basic safety standards with respect to NPP and related technologies. Recently issued new basic safety standards for radiation protection are introducing some new safety principles which may have essential impact on future licensing requirements regarding nuclear power plants and radioactive waste installations. These new standards recognize exposures under normal conditions ('practices') and intervention conditions. The term interventions describes the human activities that seek to reduce the existing radiation exposure or existing likelihood of incurring exposure which is not part of a controlled practice. The other new development in safety standards is the introduction of so called potential exposure based on the experience gained from a number of radiation accidents. This exposure is not expected to be delivered with certainty but it may result from an accident at a source or owing to an event or sequence of events of a probabilistic nature, including equipment failures and operating errors. (author)

Control Decisions for Flammable Gas Hazards in Waste Transfer Systems

International Nuclear Information System (INIS)

KRIPPS, L.J.

2000-01-01

This report describes the control decisions for flammable gas hazards in waste transfer systems (i.e., waste transfer piping and waste transfer-associated structures) made at control decision meetings on November 30, 1999a and April 19, 2000, and their basis. These control decisions, and the analyses that support them, will be documented in an amendment to the Final Safety Analysis Report (FSAR) (CHG 2000a) and Technical Safety Requirements (TSR) (CHG 2000b) to close the Flammable Gas Unreviewed Safety Question (USQ) (Bacon 1996 and Wagoner 1996). Following the Contractor Tier I review of the FSAR and TSR amendment, it will be submitted to the US. Department of Energy (DOE), Office of River Protection (ORP) for review and approval. The control decision meeting on November 30, 1999 to address flammable gas hazards in waste transfer systems followed the control decision process and the criteria for control decisions described in Section 3.3.1.5 of the FSAR. The control decision meeting agenda, attendance list, and introductory and background presentations are included in Attachments 1 through 4. The control decision discussions on existing and other possible controls for flammable gas hazards in waste transfer systems and the basis for selecting or not selecting specific controls are summarized in this report

Risk based limits for Operational Safety Requirements

International Nuclear Information System (INIS)

Cappucci, A.J. Jr.

1993-01-01

OSR limits are designed to protect the assumptions made in the facility safety analysis in order to preserve the safety envelope during facility operation. Normally, limits are set based on ''worst case conditions'' without regard to the likelihood (frequency) of a credible event occurring. In special cases where the accident analyses are based on ''time at risk'' arguments, it may be desirable to control the time at which the facility is at risk. A methodology has been developed to use OSR limits to control the source terms and the times these source terms would be available, thus controlling the acceptable risk to a nuclear process facility. The methodology defines a new term ''gram-days''. This term represents the area under a source term (inventory) vs time curve which represents the risk to the facility. Using the concept of gram-days (normalized to one year) allows the use of an accounting scheme to control the risk under the inventory vs time curve. The methodology results in at least three OSR limits: (1) control of the maximum inventory or source term, (2) control of the maximum gram-days for the period based on a source term weighted average, and (3) control of the maximum gram-days at the individual source term levels. Basing OSR limits on risk based safety analysis is feasible, and a basis for development of risk based limits is defensible. However, monitoring inventories and the frequencies required to maintain facility operation within the safety envelope may be complex and time consuming

Three-wheeled scooter taxi: A safety analysis

Indian Academy of Sciences (India)

Three-wheel scooter taxis (TSR) form an essential part of public transport for the urban ... These low cost vehicles will remain a major mode of travel in the South Asian region ... commercial codes can be avoided for computational efficiency.

Requirements to amend the main influence factors on the safety culture after fukushima accident

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.

2015-01-01

The paper presents a general model that provides a framework for the safety culture assessment, creating the possibility to identify factors that can significantly influence the safety culture. The main safety culture influence factors (SCIF) used by model are the following: regulatory environment, organizational environment, worker characteristics, socio-political environment, national culture, organization history, business and technological characteristics. After the analysis of the deficiencies and weaknesses of SCIFc in evolution of the Fukushima accident, some issues that may become necessities and requirements to change and improve both the safety culture and safety of the nuclear installations were highlighted. For each influence factor were identified some requirements to amend. The results will emphasize the necesity of the human - technology - organization system assessment. Hence it was demonstrated that the safety culture results from the interaction of individuals with technology and with the organization. (authors)

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-02-18

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

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.

General Approaches and Requirements on Safety and Security of Radioactive Materials Transport in Russian Federation

International Nuclear Information System (INIS)

Ershov, V.N.; Buchel'nikov, A.E.; Komarov, S.V.

2016-01-01

Development and implementation of safety and security requirements for transport of radioactive materials in the Russian Federation are addressed. At the outset it is worth noting that the transport safety requirements implemented are in full accordance with the IAEA's ''Regulations for the Safe Transport of Radioactive Material (2009 Edition)''. However, with respect to security requirements for radioactive material transport in some cases the Russian Federation requirements for nuclear material are more stringent compared to IAEA recommendations. The fundamental principles of safety and security of RM managements, recommended by IAEA documents (publications No. SF-1 and GOV/41/2001) are compared. Its correlation and differences concerning transport matters, the current level and the possibility of harmonization are analysed. In addition a reflection of the general approaches and concrete transport requirements is being evaluated. Problems of compliance assessment, including administrative and state control problems for safety and security provided at internal and international shipments are considered and compared. (author)

Requirements of radiation protection and safety for nuclear medicine services

International Nuclear Information System (INIS)

1989-01-01

The requirements of radiation protection and safety for nuclear medicine services are established. The norms is applied to activities related to the radiopharmaceuticals for therapeutics and 'in vivo' diagnostics purposes. (M.C.K.) [pt

Edible safety requirements and assessment standards for agricultural genetically modified organisms.

Science.gov (United States)

Deng, Pingjian; Zhou, Xiangyang; Zhou, Peng; Du, Zhong; Hou, Hongli; Yang, Dongyan; Tan, Jianjun; Wu, Xiaojin; Zhang, Jinzhou; Yang, Yongcun; Liu, Jin; Liu, Guihua; Li, Yonghong; Liu, Jianjun; Yu, Lei; Fang, Shisong; Yang, Xiaoke

2008-05-01

This paper describes the background, principles, concepts and methods of framing the technical regulation for edible safety requirement and assessment of agricultural genetically modified organisms (agri-GMOs) for Shenzhen Special Economic Zone in the People's Republic of China. It provides a set of systematic criteria for edible safety requirements and the assessment process for agri-GMOs. First, focusing on the degree of risk and impact of different agri-GMOs, we developed hazard grades for toxicity, allergenicity, anti-nutrition effects, and unintended effects and standards for the impact type of genetic manipulation. Second, for assessing edible safety, we developed indexes and standards for different hazard grades of recipient organisms, for the influence of types of genetic manipulation and hazard grades of agri-GMOs. To evaluate the applicability of these criteria and their congruency with other safety assessment systems for GMOs applied by related organizations all over the world, we selected some agri-GMOs (soybean, maize, potato, capsicum and yeast) as cases to put through our new assessment system, and compared our results with the previous assessments. It turned out that the result of each of the cases was congruent with the original assessment.

Introduction of the Amendment of IAEA Safety Requirements Reflected Lessons Learned from Fukushima Nuclear Accident

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang-Kyu; Ahn, Hyung-Joon; Kim, Sun-Hae; Cheong, Jae-Hak [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

The following five Safety Requirements publications were amended: Governmental, Legal and Regulatory Framework for Safety (GSR Part 1, 2010), Site Evaluation for Nuclear Installations (NS-R-3, 2003), Safety of Nuclear Power Plants: Design (SSR-2/1, 2012), Safety of Nuclear Power Plants: Commissioning and Operation (SSR-2/2, 2011), and Safety Assessment for Facilities and Activities (GSR Part 4, 2009). Figure 1 shows IAEA Safety Standards Categories Major amendments of five Safety Requirements publications were introduced and analyzed in this study. The five IAEA safety requirements publications which are GSR Part 1 and 4, NS-R-3 and SSR-2/1 and 2, were amended to reflect the lesson learned from the Fukushima accident and other operating experiences. Specially, 36 provisions were modified and the new 29 provision with 1 requirement (No. 67: Emergency response facilities on the site) of the SSR-2/1 were established. Since the Fukushima accident happened, a new word, design extension conditions (DECs) which cover substantially the beyond design basis accidents (BDBA), including severe accident conditions, was created and more elaborated by the world nuclear experts. Design extension conditions could include conditions in events without significant fuel degradation and conditions with core melting. Figure 2 shows the range of the DECs. The amendment of the five IAEA safety requirements publications are focused at the prevention of initiating events, which would lead to the DECs, and mitigation of the consequences of DECs by the enhanced defense in depth principle. The following examples of the IAEA requirements to prevent the initiating events are: margins for withstanding external events; margins for avoiding cliff edge effects; safety assessment for multiple facilities or activities at a single site; safety assessment in cases where resources at a facility are shared; consideration of the potential occurrence of events in combination; establishing levels of hazard

Regulations for the safe transport of radioactive material, 2005 edition. Safety requirements

International Nuclear Information System (INIS)

2005-01-01

This publication includes amendments to the 1996 Edition (As Amended 2003) arising from the second cycle of the biennial review and revision process, as agreed by the Transport Safety Standards Committee (TRANSSC) at its ninth meeting in March 2004, as endorsed by the Commission on Safety Standards at its meeting in June 2004 and as approved by the IAEA Board of Governors in November 2004. Although this publication is identified as a new edition, there are no changes that affect the administrative and approval requirements in Section VIII. The fields covered are General Provisions (radiation protection; emergency response; quality assurance; compliance assurance; non-compliance; special arrangement and training); Activity Limits and Materials Restrictions, Requirement and Controls for Transport , Requirements for Radioactive Materials and for Packagings and Packages, Test Procedures, Approval and Administrative Requirements

Safety requirements and options for a large size fast neutron reactor

International Nuclear Information System (INIS)

Cogne, F.; Megy, J.; Robert, E.; Benmergui, A.; Villeneuve, J.

1977-01-01

Starting from the experience gained in the safety evaluation of the PHENIX reactor, and from results already obtained in the safety studies on fast neutron reactors, the French regulatory bodies have defined since 1973 what could be the requirements and the recommendations in the matter of safety for the first large size ''prototype'' fast neutron power plant of 1200 MWe. Those requirements and recommendations, while not being compulsory due to the evolution of this type of reactors, will be used as a basis for the technical regulation that will be established in France in this field. They define particularly the care to be taken in the following areas which are essential for safety: the protection systems, the primary coolant system, the prevention of accidents at the core level, the measures to be taken with regard to the whole core accident and to the containment, the protection against sodium fires, and the design as a function of external aggressions. In applying these recommendations, the CREYS-MALVILLE plant designers have tried to achieve redundancy in the safety related systems and have justified the safety of the design with regard to the various involved phenomena. In particular, the extensive research made at the levels of the fuel and of the core instrumentation makes it possible to achieve the best defence to avoid the development of core accidents. The overall examination of the measures taken, from the standpoint of prevention and surveyance as well as from the standpoint of means of action led the French regulatory bodies to propose the construction permit of the CREYS MALVILLE plant, provided that additional examinations by the regulatory bodies be made during the construction of the plant on some technological aspects not fully clarified at the authorization time. The conservatism of the corresponding requirements should be demonstrated prior to the commissioning of the power plant. To pursue a programme on reactors of this type, or even more

Probabilistic safety analysis of DC power supply requirements for nuclear power plants. Technical report

International Nuclear Information System (INIS)

Baranowsky, P.W.; Kolaczkowski, A.M.; Fedele, M.A.

1981-04-01

A probabilistic safety assessment was performed as part of the Nuclear Regulatory Commission generic safety task A-30, Adequacy of Safety Related DC Power Supplies. Event and fault tree analysis techniques were used to determine the relative contribution of DC power related accident sequences to the total core damage probability due to shutdown cooling failures. It was found that a potentially large DC power contribution could be substantially reduced by augmenting the minimum design and operational requirements. Recommendations included (1) requiring DC power divisional independence, (2) improved test, maintenance, and surveillance, and (3) requiring core cooling capability be maintained following the loss of one DC power bus and a single failure in another system

Safety of transport of radioactive material. Contributed papers

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

Radioactive material has been transported for decades within and between countries as the use of radioactive material to benefit mankind has expanded. The transport can involve many types of materials (radionuclides and radiation sources for applications in agriculture, energy production, industry, and medicine) and all modes of transport (road, rail, sea and waterways, and air). Among the organizations in the United Nations system, the International Atomic Energy Agency (IAEA) has the statutory function to establish or adopt standards of safety for protection of health against exposure to ionizing radiation. Within its statutory mandate and pursuant to this request, in 1961, the IAEA issued Regulations for the Safe Transport of Radioactive Material (the IAEA Transport Regulations). The Transport Regulations were periodically reviewed and, as appropriate, have been amended or revised. The latest version of the Transport Regulations was issued in 2000 by the IAEA as Publication TS-R-1 (ST-1, Revised). In addition, the IAEA is entrusted by its Statute to provide for the application of its standards at the request of States. The objective of the Conference is to foster the exchange of information on issues related to the safety of transport of radioactive material by providing an opportunity for representatives from sponsoring international organizations and their Member States and from other co-operating and participating organizations to discuss critical issues relating to the safety of transport of radioactive material by all modes and to formulate recommendations, as appropriate, regarding further international co-operation in this area. The following topics have been identified by the Technical Programme Committee as the subjects to be covered in the background briefing sessions: History and Status of the IAEA Transport Regulation Development; Experience in adoption of the IAEA Transport Regulations at the international level; Implementation of the IAEA Transport

Safety-related requirements for photovoltaic modules and arrays

Science.gov (United States)

Levins, A.; Smoot, A.; Wagner, R.

1984-01-01

Safety requirements for photovoltaic module and panel designs and configurations for residential, intermediate, and large scale applications are investigated. Concepts for safety systems, where each system is a collection of subsystems which together address the total anticipated hazard situation, are described. Descriptions of hardware, and system usefulness and viability are included. A comparison of these systems, as against the provisions of the 1984 National Electrical Code covering photovoltaic systems is made. A discussion of the Underwriters Laboratory UL investigation of the photovoltaic module evaluated to the provisions of the proposed UL standard for plat plate photovoltaic modules and panels is included. Grounding systems, their basis and nature, and the advantages and disadvantages of each are described. The meaning of frame grounding, circuit groundings, and the type of circuit ground are covered.

75 FR 60129 - Draft Guidance for Industry and Investigators on Safety Reporting Requirements for...

Science.gov (United States)

2010-09-29

...., Bldg. 51, rm. 2201, Silver Spring, MD 20993-0002; or the Office of Communication, Outreach, and...'s ability to review critical safety information, improve safety monitoring of human drug and..., will represent the Agency's current thinking on safety reporting requirements for INDs and BA/BE...

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

1977-01-01

The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Requirements (Arabic Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication, jointly sponsored by the FAO, IAEA, ICAO, ILO, IMO, INTERPOL, OECD/NEA, PAHO, CTBTO, UNEP, OCHA, WHO and WMO, is the new edition establishing the requirements for preparedness and response for a nuclear or radiological emergency which takes into account the latest experience and developments in the area. It supersedes the previous edition of the Safety Requirements for emergency preparedness and response, Safety Standards Series No. GS-R-2, which was published in 2002. This publication establishes the requirements for ensuring an adequate level of preparedness and response for a nuclear or radiological emergency, irrespective of its cause. These Safety Requirements are intended to be used by governments, emergency response organizations, other authorities at the local, regional and national levels, operating organizations and the regulatory body as well as by relevant international organizations at the international level.

Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Requirements (Russian Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication, jointly sponsored by the FAO, IAEA, ICAO, ILO, IMO, INTERPOL, OECD/NEA, PAHO, CTBTO, UNEP, OCHA, WHO and WMO, is the new edition establishing the requirements for preparedness and response for a nuclear or radiological emergency which takes into account the latest experience and developments in the area. It supersedes the previous edition of the Safety Requirements for emergency preparedness and response, Safety Standards Series No. GS-R-2, which was published in 2002. This publication establishes the requirements for ensuring an adequate level of preparedness and response for a nuclear or radiological emergency, irrespective of its cause. These Safety Requirements are intended to be used by governments, emergency response organizations, other authorities at the local, regional and national levels, operating organizations and the regulatory body as well as by relevant international organizations at the international level.

Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Requirements (Chinese Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication, jointly sponsored by the FAO, IAEA, ICAO, ILO, IMO, INTERPOL, OECD/NEA, PAHO, CTBTO, UNEP, OCHA, WHO and WMO, is the new edition establishing the requirements for preparedness and response for a nuclear or radiological emergency which takes into account the latest experience and developments in the area. It supersedes the previous edition of the Safety Requirements for emergency preparedness and response, Safety Standards Series No. GS-R-2, which was published in 2002. This publication establishes the requirements for ensuring an adequate level of preparedness and response for a nuclear or radiological emergency, irrespective of its cause. These Safety Requirements are intended to be used by governments, emergency response organizations, other authorities at the local, regional and national levels, operating organizations and the regulatory body as well as by relevant international organizations at the international level.

Safety requirements to the operation of hydropower plants; Sicherheit beim Betrieb von Wasserkraftwerken

Energy Technology Data Exchange (ETDEWEB)

Lux, Reinhard [Berufsgenossenschaft Energie Textil Elektro Medienerzeugnisse (BG ETEM), Koeln (Germany)

2011-07-01

Employers have to take into account various safety and health requirements relating to the design, construction, operation and maintenance of hydropower plants. Especially the diversity of the hydropower plant components requires the consideration of different safety and health aspects. In 2011 the ''Fachausschuss Elektrotechnik'' (expert committee electro-technics) of the institution for statutory accident insurance and prevention presented a new ''BG-Information'' dealing with ''Safe methods operating hydropower plants''. The following article gives an introduction into the conception and the essential requirements of this new BG-Information. (orig.)

Requirements and international co-operation in nuclear safety for evolutionary light water reactors

International Nuclear Information System (INIS)

Carnino, A.

1999-01-01

The principles of safety are now well known and implemented world-wide, leading to a situation of harmonisation in accordance with the Convention on Nuclear Safety. Future reactors are expected not only to meet current requirements but to go beyond the safety level presently accepted. To this end, technical safety requirements, as defined by the IAEA document Safety Fundamentals, need be duly considered in the design, the risks to workers and population must be decreased, a stable, transparent and objective regulatory process, including an international harmonisation with respect to licensing of new reactors, must be developed, and the issue of public acceptance must be addressed. Well-performing existing installations are seen as a prerequisite for an improved public acceptability; there should be no major accidents, the results from safety performance indicators must be unquestionable, and compliance with internationally harmonised criteria is essential. Economical competitiveness is another factor that influences the acceptability; the costs for constructing the plant, for its operation and maintenance, for the fuel cycle, and for the final decommissioning are of paramount importance. Plant simplification, longer fuel cycles, life extension are appealing options, but safety will have first priority. The IAEA can play an important role in this field, by providing peer reviews by teams of international experts and assistance to Member States on the use of its safety standards. (author)

Hazard analysis & safety requirements for small drone operations : to what extent do popular drones embed safety?

NARCIS (Netherlands)

Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimichailidou, Maria Mikela

2018-01-01

Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this paper presents: (1) a set of safety

78 FR 65427 - Pipeline Safety: Reminder of Requirements for Liquefied Petroleum Gas and Utility Liquefied...

Science.gov (United States)

2013-10-31

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No. PHMSA-2013-0097] Pipeline Safety: Reminder of Requirements for Liquefied Petroleum Gas and Utility Liquefied Petroleum Gas Pipeline Systems AGENCY: Pipeline and Hazardous Materials Safety Administration...

Decommissioning of Facilities. General Safety Requirements. Pt. 6

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-15

Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning.

Combined evaluation. Plutonium transports in France. Problems of safety and reliability of transport container FS47

International Nuclear Information System (INIS)

Marignac, Y.; Coeytaux, X.; Large, J.H.

2004-09-01

This report concerns the safety and the protection of plutonium dioxide transported from Cogema La Hague to the mixed oxide fuel plant of Marcoule and Cadarache. The French approach of the transport safety is based on the combining of two essential principles: the first one affirms that the performances of the FS47 container in regard of containment (norms TS-R-1 from IAEA for the accidental conditions) is conceived to resist in any situation even terrorism or sabotage. In fact, the IAEA norm follows a probabilistic study without a voluntary attack such a terrorist one. The second principle rests on the ability to prevent the treat of terrorism acts, because of a secrecy policy on the plutonium transport. It appeared that the Green peace association has succeeded several times to know exactly the hours, the trips of the plutonium transport and this simple thing raises more questions than it solves. (N.C.)

Education and training requirements in the revised European Basic Safety Standards Directive

International Nuclear Information System (INIS)

Mundigl, S.

2009-01-01

The European Commission is currently developing a modified European Basic Safety Standards Directive covering two major objectives: the consolidation of existing European Radiation Protection legislation, and the revision of the European Basic Safety Standards. The consolidation will merge the following five Directives into one single Directive: the Basic Safety Standards Directive, the Medical Exposures Directive, the Public Information Directive, the Outside Workers Directive, and the Directive on the Control of high-activity sealed radioactive sources and orphan sources. The revision of the European Basic Safety Standards will take account of the latest recommendations by the International Commission on Radiological Protection (ICRP) and shall improve clarity of the requirements where appropriate. It is planned to introduce more binding requirements on natural radiation sources, on criteria for clearance, and on the cooperation between Member States for emergency planning and response, as well as a graded approach for regulatory control. One additional goal is to achieve greater harmonisation between the European BSS and the international BSS. Following a recommendation from the Article 31 Group of Experts, the current draft of the modified BSS will highlight the importance of education and training by dedicating a specific title to radiation protection education, training and information. This title will include a general requirement on the Member States to ensure the establishment of an adequate legislative and administrative framework for providing appropriate radiation protection education, training and information. In addition, there will be specific requirements on training in the medical field, on information and training of workers in general, of workers potentially exposed to orphan sources, and to emergency workers. The revised BSS directive will include requirements on the competence of a radiation protection expert (RPE) and of a radiation protection

A Safety and Health Guide for Vocational Educators. Incorporating Requirements of the Occupational Safety and Health Act of 1970, Relevant Pennsylvania Requirements with Particular Emphasis for Those Concerned with Cooperative Education and Work Study Programs. Volume 15. Number 1.

Science.gov (United States)

Wahl, Ray

Intended as a guide for vocational educators to incorporate the requirements of the Occupational Safety and Health Act (1970) and the requirements of various Pennsylvania safety and health regulations with their cooperative vocational programs, the first chapter of this document presents the legal implications of these safety and health…

A PLC generic requirements and specification for safety-related applications in nuclear power plants

International Nuclear Information System (INIS)

Han, Jea Bok; Lee, C. K.; Lee, D. Y.

2001-12-01

This report presents the requirements and specification to be applied to the generic qualification of programmable Logic Controller(PLC), which is being developed as part of the KNICS project, 'Development of the Digital Reactor Safety Systems' of which purpose is the application to safety-related instrumentation and control systems in nuclear power plants. This report defines the essential and critical characteristics that shall be included as part of a PLC design for safety-related application. The characteristics include performance, reliability, accuracy, the overall response time from an input to the PLC exceeding it trip condition to the resulting outputs, and the specification of processors and memories in digital controller. It also specifies the quality assurance process for software development, dealing with executive software, firmware, application software tools for developing the application software, and human machine interface(HMI). In addition, this report reviews the published standards and guidelines that are required for the PLC development and the quality assurance processes such as environment requirements, seismic withstand requirements, EMI/RFI withstand requirements, and isolation test

45 CFR 1356.30 - Safety requirements for foster care and adoptive home providers.

Science.gov (United States)

2010-10-01

... licensing file for that foster or adoptive family must contain documentation which verifies that safety... 45 Public Welfare 4 2010-10-01 2010-10-01 false Safety requirements for foster care and adoptive... ON CHILDREN, YOUTH AND FAMILIES, FOSTER CARE MAINTENANCE PAYMENTS, ADOPTION ASSISTANCE, AND CHILD AND...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Science.gov (United States)

2010-01-01

... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF...

Regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning

International Nuclear Information System (INIS)

Lipar, M.

1997-01-01

A review of regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning is given. It contains licensing steps in Slovakia during commissioning; Status and methodology of Mochovce safety analysis report; Mochovce NPP safety enhancement program; Regulatory body policy towards Mochovce NPP safety enhancement; Recent development in Mochovce pre-operational safety enhancement program review and assessment process; Licensing steps in Slovakia during commissioning

Safety requirements in the design of research reactors: A Canadian perspective

International Nuclear Information System (INIS)

Lee, A.G.; Langman, V.J.

2000-01-01

In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

Economics of the specification 6M safety re-evaluation and regulatory requirements

International Nuclear Information System (INIS)

Hopper, C.M.

1985-01-01

The objective of this work was to examine the potential economic impact of the DOT Specification 6M criticality safety re-evaluation and regulatory requirements. The examination was based upon comparative analyses of current authorized fissile material load limits for the 6M, current Federal regulations (and interpretations) limiting the contents of Type B fissile material packages, limiting aggregates of fissile material packages, and recent proposed fissile material mass limits derived from specialized criticality safety analyses of the 6M package. The work examines influences on cost in transportation, handling, and storage of fissile materials. Depending upon facility throughput requirements (and assumed incremental costs of fissile material packaging, storage, and transport), operating, facility storage capacity, and transportation costs can be reduced significantly. As an example of the pricing algorithm application based upon reasonable cost influences, the magnitude of the first year cost reductions could extend beyond four times the cost of the packaging nuclear criticality safety re-evaluation. 1 tab

Responsibility for the Violation of Ecological Safety Requirements

Science.gov (United States)

Selivanovskaya, J. I.; Gilmutdinova, I.

2018-01-01

The article deals with the problems of responsibility for the violation of ecological safety requirements from the point of view of sustainable development of the state. Such types of responsibility as property, disciplinary, financial, administrative and criminal responsibility in the area are analysed. Suggestions on the improvement of legislation are put forward. Among other things it is suggested to introduce criminal sanctions against legal bodies (enterprises) for ecological crimes with punishments in the form of fines, suspension or discontinuation of activities.

RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

International Nuclear Information System (INIS)

KOZLOWSKI, S.D.

2007-01-01

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

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)

Safety requirements for a nuclear power plant electric power system

Energy Technology Data Exchange (ETDEWEB)

Fouad, L F; Shinaishin, M A

1988-06-15

This work aims at identifying the safety requirements for the electric power system in a typical nuclear power plant, in view of the UNSRC and the IAEA. Description of a typical system is provided, followed by a presentation of the scope of the information required for safety evaluation of the system design and performance. The acceptance and design criteria that must be met as being specified by both regulatory systems, are compared. Means of implementation of such criteria as being described in the USNRC regulatory guides and branch technical positions on one hand and in the IAEA safety guides on the other hand are investigated. It is concluded that the IAEA regulations address the problems that may be faced with in countries having varying grid sizes ranging from large stable to small potentially unstable ones; and that they put emphasis on the onsite standby power supply. Also, in this respect the Americans identify the grid as the preferred power supply to the plant auxiliaries, while the IAEA leaves the possibility that the preferred power supply could be either the grid or the unit main generator depending on the reliability of each. Therefore, it is found that it is particularly necessary in this area of electric power supplies to deal with the IAEA and the American sets of regulations as if each complements and not supplements the other. (author)

User requirements in the area of safety of innovative nuclear reactors and fuel cycle installations

International Nuclear Information System (INIS)

Kuczera, B.; Juhn, P.E.; Fukuda, K.; )

2002-01-01

Full text: Against the background of already existing IAEA and INSAC publications in the area of safety, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a set of user requirements for the safety of future nuclear installations has been established. Five top-level requirements are expected to apply to any type of innovative design. They should foster an increased level of safety that is transparent to and fully accepted by the general public. The approach to future reactor safety includes two complementary strategies: increased emphasis on inherent safety characteristics and enhancement of defense in depth. As compared to existing plants, the effectiveness of preventing measures should be highly enhanced, resulting in fewer mitigation measures. The targets and possible approaches of each of the five levels of defense developed for innovative reactor designs are outlined in the paper

78 FR 42889 - Pipeline Safety: Reminder of Requirements for Utility LP-Gas and LPG Pipeline Systems

Science.gov (United States)

2013-07-18

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part 192 [Docket No. PHMSA-2013-0097] Pipeline Safety: Reminder of Requirements for Utility LP-Gas and LPG Pipeline Systems AGENCY: Pipeline and Hazardous Materials Safety Administration (PHMSA), DOT. ACTION...

Safety research needs for Russian-designed reactors. Requirements situation

International Nuclear Information System (INIS)

Brown, R. Allan; Holmstrom, Heikki; Reocreux, Michel; Schulz, Helmut; Liesch, Klaus; Santarossa, Giampiero; Hayamizu, Yoshitaka; Asmolov, Vladimir; Bolshov, Leonid; Strizhov, Valerii; Bougaenko, Sergei; Nikitin, Yuri N.; Proklov, Vladimir; Potapov, Alexandre; Kinnersly, Stephen R.; Voronin, Leonid M.; Honekamp, John R.; Frescura, Gianni M.; Maki, Nobuo; Reig, Javier; ); Bekjord, Eric S.; Rosinger, Herbert E.

1998-01-01

integrity must be verified, and material property data bases extended. - VVER severe accident research should focus on validation of codes for accident management procedures, and on extension and qualification of an appropriate data base for materials properties and their interactions. - RBMK thermal-hydraulic research is needed to improve the technical basis for further development of RBMK safety criteria. - Assessment of the integrity of the RBMK primary coolant circuit, and especially the fuel channel, requires urgent research. Methods of assessing RBMK pressure boundary integrity must be verified, and material property data bases extended. - RBMK severe accident research should focus on prevention of accidents and Accident Management for cases of loss of heat sink and Beyond Design-Basis Loss-of-Coolant Accidents. For these purposes, simple physical models and parametric codes need development and should be systematically used in plant specific analysis. Recommendations; - A Safety Research Strategic Plan should be developed. Such a plan sets goals, defines products, and describes when and how work will be done, including determination of research priorities. - Key players, including regulators, operators, plant designers and researchers should be involved in developing and implementing this plan and its execution and applying the results. - International cooperation in safety research should be encouraged for purposes of improving quality, preventing technical isolation and cost sharing. - New approaches, such as technical fora for specific technical topics, should be established to make safety research information in OECD countries available to researchers working on the safety of Russian-designed reactors

Proposal for basic safety requirements regarding the disposal of high-level radioactive waste

International Nuclear Information System (INIS)

1980-04-01

A working group commissioned to prepare proposals for basic safety requirements for the storage and transport of radioactive waste prepared its report to the Danish Agency of Environmental Protection. The proposals include: radiation protection requirements, requirements concerning the properties of high-level waste units, the geological conditions of the waste disposal location, the supervision of waste disposal areas. The proposed primary requirements for safety evaluation of the disposal of high-level waste in deep geological formations are of a general nature, not being tied to specific assumptions regarding the waste itself, the geological and other conditions at the place of disposal, and the technical methods of disposal. It was impossible to test the proposals for requirements on a working repository. As no country has, to the knowledge of the working group, actually disposed of hifg-level radioactive waste or approved of plans for such disposal. Methods for evaluating the suitability of geological formations for waste disposal, and background material concerning the preparation of these proposals for basic safety requirements relating to radiation, waste handling and geological conditions are reviewed. Appended to the report is a description of the phases of the fuel cycle that are related to the storage of spent fuel and the disposal of high-level reprocessing waste in a salt formation. It should be noted that the proposals of the working group are not limited to the disposal of reprocessed fuel, but also include the direct disposal of spent fuel as well as disposal in geological formations other than salt. (EG)

Evaluation of safety requirements of erbium laser equipment used in dentistry

International Nuclear Information System (INIS)

Braga, Flavio Hamilton

2002-01-01

The erbium laser (Er:YAG) has been used in several therapeutic processes. Erbium lasers, however, operate with energies capable to produce lesions in biological tissues. Aiming the safe use, the commercialization of therapeutic laser equipment is controlled in Brazil, where the equipment should comply with quality and safety requirement prescribed in technical regulations. The objective of this work is to evaluate the quality and safety requirements of a commercial therapeutic erbium laser according to Brazilian regulations, and to discuss a risk control program intended to minimize the accidental exposition at dangerous laser radiation levels. It was verified that the analyzed laser can produce lesions in the skin and eyes, when exposed to laser radiation at distances smaller than 80 cm by 10 s or more. In these conditions, the use of protection glasses is recommended to the personnel that have access to the laser operation ambient. It was verified that the user's training and the presence of a target indicator are fundamental to avoid damages in the skin and buccal cavity. It was also verified that the knowledge and the correct use of the equipment safety devices, and the application of technical and administrative measures is efficient to minimize the risk of dangerous expositions to the laser radiation. (author)

Statement on safety requirements concerning the long-term operation of the Muehleberg nuclear power station

International Nuclear Information System (INIS)

2012-12-01

This report published by the Swiss Federal Nuclear Safety Inspectorate ENSI investigates the safety requirements with respect to the long-term operation of the Muehleberg nuclear power station in Switzerland. Relevant international requirements and Swiss legal stipulations concerning the long-term operation of the power station are stated. The management of aging processes is looked at. The regular verification of the integrity of various plant components such as containments, piping, steam generation system, etc. is looked at in detail. The state-of-the-art concerning deterministic accident analyses and refitting technology are discussed, as are automated safety systems. The applicable laws, decrees and guidelines are listed in appendices

Disposal of Radioactive Waste. Specific Safety Requirements

International Nuclear Information System (INIS)

2011-01-01

The IAEA's Statute authorizes the Agency to 'establish or adopt... standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the

Implications of safety requirements for the treatment of THMC processes in geological disposal systems for radioactive waste

Directory of Open Access Journals (Sweden)

Frédéric Bernier

2017-06-01

Full Text Available The mission of nuclear safety authorities in national radioactive waste disposal programmes is to ensure that people and the environment are protected against the hazards of ionising radiations emitted by the waste. It implies the establishment of safety requirements and the oversight of the activities of the waste management organisation in charge of implementing the programme. In Belgium, the safety requirements for geological disposal rest on the following principles: defence-in-depth, demonstrability and the radiation protection principles elaborated by the International Commission on Radiological Protection (ICRP. Applying these principles requires notably an appropriate identification and characterisation of the processes upon which the safety functions fulfilled by the disposal system rely and of the processes that may affect the system performance. Therefore, research and development (R&D on safety-relevant thermo-hydro-mechanical-chemical (THMC issues is important to build confidence in the safety assessment. This paper points out the key THMC processes that might influence radionuclide transport in a disposal system and its surrounding environment, considering the dynamic nature of these processes. Their nature and significance are expected to change according to prevailing internal and external conditions, which evolve from the repository construction phase to the whole heating–cooling cycle of decaying waste after closure. As these processes have a potential impact on safety, it is essential to identify and to understand them properly when developing a disposal concept to ensure compliance with relevant safety requirements. In particular, the investigation of THMC processes is needed to manage uncertainties. This includes the identification and characterisation of uncertainties as well as for the understanding of their safety-relevance. R&D may also be necessary to reduce uncertainties of which the magnitude does not allow

Requirements of safety and reliability

International Nuclear Information System (INIS)

Franzen, L.F.

1977-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the findings derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essential for accident analyses, and the determination of the loads occuring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig./HP) [de

Development of safety-related regulatory requirements for nuclear power in developing countries. Key issue paper no. 4

International Nuclear Information System (INIS)

Han, K.I.

2000-01-01

In implementing a national nuclear power program, balanced regulatory requirements are necessary to ensure nuclear safety and cost competitive nuclear power, and to help gain public acceptance. However, this is difficult due to the technology-intensive nature of the nuclear regulatory requirements, the need to reflect evolving technology and the need for cooperation among multidisciplinary technical groups. This paper suggests approaches to development of balanced nuclear regulatory requirements in developing countries related to nuclear power plant safety, radiation protection and radioactive waste management along with key technical regulatory issues. It does not deal with economic or market regulation of electric utilities using nuclear power. It suggests that national regulatory requirements be developed using IAEA safety recommendations as guidelines and safety requirements of the supplier country as a main reference after careful planning, manpower buildup and thorough study of international and supplier country's regulations. Regulation making is not recommended before experienced manpower has been accumulated. With an option that the supplier country's regulations may be used in the interim, the lack of complete national regulatory requirements should not deter introduction of nuclear power in developing countries. (author)

76 FR 64 - Safety and Health Requirements Related to Camp Cars

Science.gov (United States)

2011-01-03

.... Water uses such as personal oral hygiene, drinking, food washing, preparation, cooking, cleaning of the... of Sec. 228.325 to ensure that the food service is safe and sanitary. FRA will hold the railroad... proposed section sets forth requirements regarding the safety of heating, cooking, ventilation, air...

A Review of Safety and Design Requirements of the Artificial Pancreas

NARCIS (Netherlands)

Blauw, Helga; Keith-Hynes, Patrick; Koops, Robin; DeVries, J. Hans

2016-01-01

As clinical studies with artificial pancreas systems for automated blood glucose control in patients with type 1 diabetes move to unsupervised real-life settings, product development will be a focus of companies over the coming years. Directions or requirements regarding safety in the design of an

Relationship between general safety requirements and safety culture in the improvement of safe operation of I.N.R. TRIGA reactor facilities

International Nuclear Information System (INIS)

Ciocanescu, M.; Preda, M.; Chiritescu, M.; Dumitru, M.

1996-01-01

Acquiring of the basic principles of ''safety culture'' by a large number of profesionals in the nuclear field drew the attention of the decision factors in the INR managerial structure, who decided to promote certain practical actions at each level in order to improve nuclear safety. Starting from the ''Republican Standards for Nuclear Safety'' issued by CSEN in 1975, where general safety criteria are defined for nuclear reactors and NPPs, the specialists at the TRIGA reactor originated and implemented a coherent and secure system to ensure nuclear safety over all steps of nuclear activities: research, conception, execution, commissioning and operation. This system has been continuosly corrected so that now it is completely integrated in a modern safety system. The paper presents the way in which a modern system for nuclear safety at the TRIGA reactor has been implemented and developed, in accordance to specific criteria and requirements imposed by related National Regulations and with the principles of safety culture. Starting from the definition of specific responsabilities, there are presented the internal stipulations and practical actions at all levels in order to enhance nuclear safety. (orig.)

Safety requirements and feedback of commonly used material handling equipment

International Nuclear Information System (INIS)

Pathak, M.K.

2009-01-01

Different types of cranes, hoists, chain pulley blocks are the most commonly used material handling equipment in industry along with attachments like chains, wire rope slings, d-shackles, etc. These equipment are used at work for transferring loads from one place to another and attachments are used for anchoring, fixing or supporting the load. Selection of the correct equipment, identification of the equipment planning of material handling operation, examination/testing of the equipment, education and training of the persons engaged in operation of the material handling equipment can reduce the risks to safety of people in workplace. Different safety systems like boom angle indicator, overload tripping device, limit switches, etc. should be available in the cranes for their safe use. Safety requirement for safe operation of material handling equipment with emphasis on different cranes and attachments particularly wire rope slings and chain slings have been brought out in this paper. An attempt has also been made to bring out common nature of deficiencies observed during regulatory inspection carried out by AERB. (author)

JET-ISX-B beryllium limiter experiment safety analysis report and operational safety requirements

International Nuclear Information System (INIS)

Edmonds, P.H.

1985-09-01

An experiment to evaluate the suitability of beryllium as a limiter material has been completed on the ISX-B tokamak. The experiment consisted of two phases: (1) the initial operation and characterization in the ISX experiment, and a period of continued operation to the specified surface fluence (10 22 atoms/cm 2 ) of hydrogen ions; and (2) the disassembly, decontamination, or disposal of the ISX facility. During these two phases of the project, the possibility existed for beryllium and/or beryllium oxide powder to be produced inside the vacuum vessel. Beryllium dust is a highly toxic material, and extensive precautions are required to prevent the release of the beryllium into the experimental work area and to prevent the contamination of personnel working on the device. Details of the health hazards associated with beryllium and the appropriate precautions are presented. Also described in appendixes to this report are the various operational safety requirements for the project

Eurosafe 2006 radioactive waste management: long term safety requirements and societal expectations

International Nuclear Information System (INIS)

2006-01-01

The EUROSAFE Forum is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety, to share experiences, exchange technical and scientific opinions, and conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum 2006 focuses on 'Radioactive Waste Management: Long Term Safety Requirements and Societal Expectations' from the point of view of the authorities, TSOs and industry and presents the latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe. A high level of nuclear safety is a priority for Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining increasing recognition. This

Eurosafe 2006 radioactive waste management: long term safety requirements and societal expectations

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

The EUROSAFE Forum is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety, to share experiences, exchange technical and scientific opinions, and conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum 2006 focuses on 'Radioactive Waste Management: Long Term Safety Requirements and Societal Expectations' from the point of view of the authorities, TSOs and industry and presents the latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe. A high level of nuclear safety is a priority for Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining increasing recognition. This

Eurosafe 2006 radioactive waste management: long term safety requirements and societal expectations

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

The EUROSAFE Forum is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety, to share experiences, exchange technical and scientific opinions, and conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum 2006 focuses on 'Radioactive Waste Management: Long Term Safety Requirements and Societal Expectations' from the point of view of the authorities, TSOs and industry and presents the latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe. A high level of nuclear safety is a priority for Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining increasing recognition

Construction products performances and basic requirements for fire safety of facades in energy rehabilitation of buildings

Directory of Open Access Journals (Sweden)

Laban Mirjana Đ.

2015-01-01

Full Text Available Construction product means any product or kit which is produced and placed on the market for incorporation in a permanent manner in construction works, or parts thereof, and the performance of which has an effect on the performance of the construction works with respect to the basic requirements for construction works. Safety in case of fire and Energy economy and heat retention represent two among seven basic requirements which building has to meet according to contemporary technical rules on planning and construction. Performances of external walls building materials (particularly reaction to fire could significantly affect to fire spread on the façade and other building parts. Therefore, façade shaping and materialization in building renewal process, has to meet the fire safety requirement, as well as the energy requirement. Brief survey of fire protection regulations development in Serbia is presented in the paper. Preventive measures for fire risk reduction in building façade energy renewal are proposed according to contemporary fire safety requirements.

Safety Culture: A Requirement for New Business Models — Lessons Learned from Other High Risk Industries

International Nuclear Information System (INIS)

Kecklund, L.

2016-01-01

Technical development and changes on global markets affects all high risk industries creating opportunities as well as risks related to the achievement of safety and business goals. Changes in legal and regulatory frameworks as well as in market demands create a need for major changes. Several high risk industries are facing a situation where they have to develop new business models. Within the transportation domain, e.g., aviation and railways, there is a growing concern related to how the new business models may affects safety issues. New business models in aviation and railways include extensive use of outsourcing and subcontractors to reduce costs resulting in, e.g., negative changes in working conditions, work hours, employment conditions and high turnover rates. The energy sector also faces pressures to create new business models for transition to renewable energy production to comply with new legal and regulatory requirements and to make best use of new reactor designs. In addition, large scale phase out and decommissioning of nuclear facilities have to be managed by the nuclear industry. Some negative effects of new business models have already arisen within the transportation domain, e.g., the negative effects of extensive outsourcing and subcontractor use. In the railway domain the infrastructure manager is required by European and national regulations to assure that all subcontractors are working according to the requirements in the infrastructure managers SMS (Safety Management System). More than ten levels of subcontracts can be working in a major infrastructure project making the system highly complex and thus difficult to control. In the aviation domain, tightly coupled interacting computer networks supplying airport services, as well as air traffic control, are managed and maintained by several different companies creating numerous interfaces which must be managed by the SMS. There are examples where a business model with several low

Software Safety Risk in Legacy Safety-Critical Computer Systems

Science.gov (United States)

Hill, Janice L.; Baggs, Rhoda

2007-01-01

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

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

Modeling the Non-functional Requirements in the Context of Usability, Performance, Safety and Security

OpenAIRE

Sadiq, Mazhar

2007-01-01

Requirement engineering is the most significant part of the software development life cycle. Until now great emphasis has been put on the maturity of the functional requirements. But with the passage of time it reveals that the success of software development does not only pertain to the functional requirements rather non-functional requirements should also be taken into consideration. Among the non-functional requirements usability, performance, safety and security are considered important. ...

78 FR 55230 - Safety and Environmental Management System Requirements for Vessels on the U.S. Outer Continental...

Science.gov (United States)

2013-09-10

...\\ including the regulation of workplace safety and health.\\2\\ The Coast Guard's regulatory authority extends... 147 [Docket No. USCG-2012-0779] RIN 1625-AC05 Safety and Environmental Management System Requirements... a vessel-specific Safety and Environmental Management System (SEMS) that incorporates the management...

Evaluation and qualification of novel control techniques with safety requirements

International Nuclear Information System (INIS)

Gossner, S.; Wach, D.

1985-01-01

The paper discusses the questions related to the assessment and qualification of new I and C-systems. The tasks of nuclear power plant I and Cs as well as the efficiency of the new techniques are reflected. Problems with application of new I and Cs and the state of application in Germany and abroad are addressed. Starting from the essential differencies between conventional and new I and C-systems it is evaluated, if and in which way existing safety requirements can be met and to what extent new requirements need to be formulated. An overall concept has to be developed comprising the definition of graded requirement profiles for design and qualification. Associated qualification procedures and tools have to be adapted, developed and tuned upon each other. (orig./HP) [de

Probabilistic approaches to LCO's and surveillance requirements for standby safety systems

International Nuclear Information System (INIS)

Lofgren, E.V.; Varcolik, F.

1982-11-01

Results are presented for a comprehensive analysis of risk-based methods for establishing Limiting Conditions for Operation (LCO) and surveillance requirements for on-line test and repair of nuclear power plant safety system components. Limiting Conditions for Operation refers to the legal constraint on safety system component outage times that are imposed by the NRC as part of the reactor operating license. Generally, when a safety system component is removed for repair or test for a period of time there is a period of increased vulnerability concerning the probability that the affected safety system will be available to mitigate an accident. This period of increased vulnerability exists until the component is restored to service. The constraint on the duration of this period, the allowed outage time (AOT), is the aspect of LCOs that is of interest here. In particular, methods are reviewed and developed that relate measures of risk to the AOT. Only by explicitly relating risk to AOT can outage times be constrained by placing limits on risk. Methods developed for relating risk measures to outage times are presented. The review and analysis of risk related methods for establishing LCOs are described

Identifying environmental safety and health requirements for an Environmental Restoration Management Contractor

International Nuclear Information System (INIS)

Beckman, W.H.; Cossel, S.C.; Alhadeff, N.; Lindamood, S.B.; Beers, J.A.

1993-10-01

The purpose of the Standards/Requirements Identification Program, developed partially in response to the Defense Nuclear Facilities Safety Board Recommendation 90-2, was to identify applicable requirements that established the Environmental Restoration Management Contractor's (ERMC) responsibilities and authorities under the Environmental Restoration Management Contract, determine the adequacy of these requirements, ascertain a baseline level of compliance with them, and implement a maintenance program that would keep the program current as requirements or compliance levels change. The resultant Standards/Requirements Identification Documents (S/RIDs) consolidate the applicable requirements. These documents govern the development of procedures and manuals to ensure compliance with the requirements. Twenty-four such documents, corresponding with each functional area identified at the site, are to be issued. These requirements are included in the contractor's management plan

Regulatory Safety Requirements for Operating Nuclear Installations

International Nuclear Information System (INIS)

Gubela, W.

2017-01-01

The National Nuclear Regulator (NNR) is established in terms of the National Nuclear Regulator Act (Act No 47 of 1999) and its mandate and authority are conferred through sections 5 and 7 of this Act, setting out the NNR's objectives and functions, which include exercising regulatory control over siting, design, construction etc of nuclear installations through the granting of nuclear authorisations. The NNR's responsibilities embrace all those actions aimed at providing the public with confidence and assurance that the risks arising from the production of nuclear energy remain within acceptable safety limits -> Therefore: Set fundamental safety standards, conducting pro-active safety assessments, determining licence conditions and obtaining assurance of compliance. The promotional aspects of nuclear activities in South Africa are legislated by the Nuclear Energy Act (Act No 46 of 1999). The NNR approach to regulations of nuclear safety and security take into consideration, amongst others, the potential hazards associated with the facility or activity, safety related programmes, the importance of the authorisation holder's safety related processes as well as the need to exercise regulatory control over the technical aspects such as of the design and operation of a nuclear facility in ensuring nuclear safety and security. South Africa does not have national nuclear industry codes and standards. The NNR is therefore non-prescriptive as it comes to the use of industry codes and standards. Regulatory framework (current) provide for the protection of persons, property, and environment against nuclear damage, through Licensing Process: Safety standards; Safety assessment; Authorisation and conditions of authorisation; Public participation process; Compliance assurance; Enforcement

Requirements of radiation and safety protection for NORM in petroleum and gas facilities

International Nuclear Information System (INIS)

Machavane, Edna Felicina Lisboa

2017-01-01

The work establishes radiation protection and safety requirements for NORM in oil and gas installations, enabling the National Atomic Energy Agency to draw up regulations on NORM. A bibliographic review and measurement of oil sludge activity concentrations was carried out to reach the objective. Significant amounts of NORM originating from reservoir rock are encountered during production, maintenance and decommissioning. The oil and gas industry operates in all climates and environments including the most arduous conditions and is continually challenged to achieve high operating efficiency while maintaining a high standard of safety and control - this includes the need to maintain control over exposure as well as protecting the public and the environment through the proper management of tailings that may be radiologically and chemically hazardous. The main objective of this work was not only to present the main radiological protection and safety requirements for NORM in oil and gas installations, but also to guide the competent governmental authorities of the Republic of Mozambique, that the installation of a radiometry laboratory and elaboration of NORM regulations involve a great control of radiological safety. The regulatory authority is responsible for authorizing facilities for the storage of radioactive waste, including the storage of contaminated tailings. It is recommended that studies of this kind be made to analyze the concentration of naturally occurring radioisotope activity. (author)

Decommissioning of Facilities. General Safety Requirements. Pt. 6 (Spanish Edition)

International Nuclear Information System (INIS)

2017-01-01

Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning.

Decommissioning of Facilities. General Safety Requirements. Pt. 6 (Russian Edition)

International Nuclear Information System (INIS)

2015-01-01

Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning

Software safety analysis on the model specified by NuSCR and SMV input language at requirements phase of software development life cycle using SMV

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun

2005-01-01

Safety-critical software process is composed of development process, verification and validation (V and V) process and safety analysis process. Safety analysis process has been often treated as an additional process and not found in a conventional software process. But software safety analysis (SSA) is required if software is applied to a safety system, and the SSA shall be performed independently for the safety software through software development life cycle (SDLC). Of all the phases in software development, requirements engineering is generally considered to play the most critical role in determining the overall software quality. NASA data demonstrate that nearly 75% of failures found in operational software were caused by errors in the requirements. The verification process in requirements phase checks the correctness of software requirements specification, and the safety analysis process analyzes the safety-related properties in detail. In this paper, the method for safety analysis at requirements phase of software development life cycle using symbolic model verifier (SMV) is proposed. Hazard is discovered by hazard analysis and in other to use SMV for the safety analysis, the safety-related properties are expressed by computation tree logic (CTL)

International review on safety requirements for the prototype fast breeder reactor “Monju”

International Nuclear Information System (INIS)

2016-01-01

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

Safety requirements and radiological protection for ore installations

International Nuclear Information System (INIS)

2003-06-01

This norm establishes the safety and radiological protection requirements for mining installations which manipulates, process and storing ores, raw materials, steriles, slags and wastes containing radionuclides of the uranium and thorium natural series, simultaneously or separated, and which can cause undue exposures to the public and workers, at anytime of the functioning or pos operational stage. This norm applies to the mining installations activities, suspended or which have ceased their activities before the issue date of this norm, destined to the mining, physical, chemical and metallurgical processing, and the industrialization of raw materials and residues containing associated radionuclides from the natural series of uranium and thorium, including the stages of implantation, operation and decommissioning of the installation

78 FR 47015 - Software Requirement Specifications for Digital Computer Software Used in Safety Systems of...

Science.gov (United States)

2013-08-02

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0195] Software Requirement Specifications for Digital Computer Software Used in Safety Systems of Nuclear Power Plants AGENCY: Nuclear Regulatory Commission... issuing a revised regulatory guide (RG), revision 1 of RG 1.172, ``Software Requirement Specifications for...

[Storage of plant protection products in farms: minimum safety requirements].

Science.gov (United States)

Dutto, Moreno; Alfonzo, Santo; Rubbiani, Maristella

2012-01-01

Failure to comply with requirements for proper storage and use of pesticides in farms can be extremely hazardous and the risk of accidents involving farm workers, other persons and even animals is high. There are still wide differences in the interpretation of the concept of "securing or making safe", by workers in this sector. One of the critical points detected, particularly in the fruit sector, is the establishment of an adequate storage site for plant protection products. The definition of "safe storage of pesticides" is still unclear despite the recent enactment of Legislative Decree 81/2008 regulating health and work safety in Italy. In addition, there are no national guidelines setting clear minimum criteria for storage of plant protection products in farms. The authors, on the basis of their professional experience and through analysis of recent legislation, establish certain minimum safety standards for storage of pesticides in farms.

Safety evaluations required in the safety regulations for Monju and the validity confirmation of safety evaluation methods

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purposes of this study are to perform the safety evaluations of the fast breeder reactor 'Monju' and to confirm the validity of the safety evaluation methods. In JFY 2012, the following results were obtained. As for the development of safety evaluation methods needed in the safety examination achieved for the reactor establishment permission, development of the analysis codes, such as a core damage analysis code, were carried out according to the plan. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

78 FR 2797 - Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and Electric Vehicles

Science.gov (United States)

2013-01-14

... Sound Requirements for Hybrid and Electric Vehicles; Draft Environmental Assessment for Rulemaking To Establish Minimum Sound Requirements for Hybrid and Electric Vehicles; Proposed Rules #0;#0;Federal Register...-0148] RIN 2127-AK93 Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and...

Nuclear safety requirements for upgrading the National Repository for Radioactive Wastes-Baita Bihor

International Nuclear Information System (INIS)

Vladescu, Gabriela; Necula, Daniela

2000-01-01

The upgrading project of National Repository for Radioactive Wastes-Baita Bihor is based on the integrated concept of nuclear safety. Its ingredients are the following: A. The principles of nuclear safety regarding the management of radioactive wastes and radioprotection; B. Safety objectives for final disposal of low- and intermediate-level radioactive wastes; C. Safety criteria for final disposal of low- and intermediate-level radioactive wastes; D. Assessment of safety criteria fulfillment for final disposal of low- and intermediate-level radioactive wastes. Concerning the nuclear safety in radioactive waste management the following issues are considered: population health protection, preventing transfrontier contamination, future generation radiation protection, national legislation, control of radioactive waste production, interplay between radioactive waste production and management, radioactive waste repository safety. The safety criteria of final disposal of low- and intermediate-level radioactive wastes are discussed by taking into account the geological and hydrogeological configuration, the physico-chemical and geochemical characteristics, the tectonics and seismicity conditions, extreme climatic potential events at the mine location. Concerning the requirements upon the repository, the following aspects are analyzed: the impact on environment, the safety system reliability, the criticality control, the filling composition to prevent radioactive leakage, the repository final sealing, the surveillance. Concerning the radioactive waste, specific criteria taken into account are the radionuclide content, the chemical composition and stability, waste material endurance to heat and radiation. The waste packaging criteria discussed are the mechanical endurance, materials toughness and types as related to deterioration caused by handling, transportation, storing or accidents. Fulfillment of safety criteria is assessed by scenarios analyses and analyses of

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

International Nuclear Information System (INIS)

Franzen, L.F.

1977-01-01

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

IAEA Safety Standards

International Nuclear Information System (INIS)

2016-09-01

The IAEA Safety Standards Series comprises publications of a regulatory nature covering nuclear safety, radiation protection, radioactive waste management, the transport of radioactive material, the safety of nuclear fuel cycle facilities and management systems. These publications are issued under the terms of Article III of the IAEA’s Statute, which authorizes the IAEA to establish “standards of safety for protection of health and minimization of danger to life and property”. Safety standards are categorized into: • Safety Fundamentals, stating the basic objective, concepts and principles of safety; • Safety Requirements, establishing the requirements that must be fulfilled to ensure safety; and • Safety Guides, recommending measures for complying with these requirements for safety. For numbering purposes, the IAEA Safety Standards Series is subdivided into General Safety Requirements and General Safety Guides (GSR and GSG), which are applicable to all types of facilities and activities, and Specific Safety Requirements and Specific Safety Guides (SSR and SSG), which are for application in particular thematic areas. This booklet lists all current IAEA Safety Standards, including those forthcoming

Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

Energy Technology Data Exchange (ETDEWEB)

Nonaka, N.; Kawatta, N.; Niwa, H.; Kondo, S.; Maeda, K

1996-12-31

This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given.

What Isn't Working and New Requirements. The Need to Harmonize Safety and Security Requirements

International Nuclear Information System (INIS)

Flory, D.

2011-01-01

The year 2011 marks the 50th anniversary of the first IAEA regulations governing the transport of radioactive material. However transport safety at the IAEA obviously predates this, since the regulations took time to develop. In 1957, GC. 1/1 already states: 'The Agency should undertake studies with a view to the establishment of regulations relating to the international transportation of radioactive materials. ...'. And goes further: 'The transport of radioisotopes and radiation sources has brought to light many problems and involves the need for uniform packaging and shipping regulations ... facilitate the acceptance of such materials by sea and air carriers'. This conference reiterates the challenge given then through the sub-title 'The next fifty years - Creating a Safe, Secure and Sustainable Framework'. Looking back, we can see that the sustainable framework was a goal in 1957, where radioactive material could be transported should it be desired. Since these early days we have added to safety the need to ensure security. However we still see the same calls today to eradicate denial of shipment, which might suggest we have not progressed. But the picture today is very different - we have today well established requirements for safe transport of radioactive material, and the recommendations for security in transport are coming of age for all radioactive materials. The outstanding issue would seem to be harmonisation, not just between safety and security in IAEA documents, but also harmonisation between Member States.

Quantitative Comparison of the Walk and Trot of Border Collies and Labrador Retrievers, Breeds with Different Performance Requirements

Science.gov (United States)

Carr, Brittany Jean; Canapp, Sherman O; Zink, M. Christine

2015-01-01

Introduction We hypothesized that breed differences of Border Collies and Labrador Retrievers would be reflected in the temporospatial characteristics of the walk and trot. Materials and Methods Twenty healthy Border Collies and 20 healthy Labrador Retrievers made three passes across a pressure sensing walkway system that recorded quantitative temporospatial information at a walk and a trot. The following variables were measured for each dog: velocity, total pressure index percentage (TPI%), ratio of weight borne on the thoracic vs. pelvic limbs (T/P TPI%), stance time percentage (ST%), and thoracic limb stride length (TSrL). Results The mean T/P TPI% for Border Collies at a walk and at a trot were significantly lower than for Labrador Retrievers (p = 0.0007 and p = 0.0003). Border Collies had a significantly lower ST% than Labrador Retrievers for the thoracic limbs and pelvic limbs at a walk (p = 0.0058 and 0.0003) and the trot (p = 0.0280 and 0.0448). There was no relationship between ST% and TSrL in Border Collies and an inverse correlation between ST% and TSrL in Labrador Retrievers (p = 0.0002). Discussion Key quantitative gait differences were identified in Border Collies and Labrador Retrievers, which could potentially provide each breed with an advantage for their working function. PMID:26689372

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

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

Qualification testing facility for packages to be used for transport and storage of radioactive materials

International Nuclear Information System (INIS)

Vieru, Gheorghe

2009-01-01

The radioactive materials (RAM) packaging have to comply to all modes and transport condition, routine or in accident conditions possibly to occur during transportation operations. It is well known that the safety in the transport of RAM is dependent on packaging appropriate for the contents being shipped rather than on operational and/or administrative actions required for the package. The quality of these packages - type A, B or C has to be proved by performing qualification tests in accordance with the ROMANIAN nuclear regulation conditions provided by CNCAN Order no. 357/22.12.2005- 'Norms for a Safe Transport of Radioactive Material', the IAEA Vienna Recommendation stipulated in the Safety standard TS-R-1- Regulation for the Safe Transport of Radioactive Material, 2005 Edition, and other applicable international recommendations. The paper will describe the components of the designed testing facilities, and the qualification testing to be performed for all type A, B and C packages subjected to the testing. In addition, a part of the qualification tests for a package (designed and manufactured in INR Pitesti) used for transport and storage of spent fuel LEU elements of a TRIGA nuclear reactor will be described and analyzed. Quality assurance and quality controls measures taken in order to meet technical specification provided by the design are also presented and commented. The paper concludes that the new Romanian Testing Facilities for RAM packages will comply with the national safe standards as well as with the IAEA applicable recommendation provided by the TS-R-1 safety standard. (author)

Proceedings for the Advance Planning Briefing for Industry

Science.gov (United States)

1990-01-24

Liaison Office TOD - Technical Objective Documents TSR - Tactical Source Region UAV - Unmanned Aerial Vehicle UGT - UnderGround nuclear Test A G.EN D635I...tests in AURORA and underground nuclear tests ( UGT ) and will help develop tactical source region hardening requirements and lead to approaches for TSR...X-Ray theory , lasers, electronic controllers, computers, robotics, etc. Contracting for scientific studies and one-of-a-kind machines will emphasize

Risk and safety requirements for diagnostic and therapeutic procedures in allergology: World Allergy Organization Statement

Directory of Open Access Journals (Sweden)

Marek L. Kowalski

2016-10-01

Full Text Available Abstract One of the major concerns in the practice of allergy is related to the safety of procedures for the diagnosis and treatment of allergic disease. Management (diagnosis and treatment of hypersensitivity disorders involves often intentional exposure to potentially allergenic substances (during skin testing, deliberate induction in the office of allergic symptoms to offending compounds (provocation tests or intentional application of potentially dangerous substances (allergy vaccine to sensitized patients. These situations may be associated with a significant risk of unwanted, excessive or even dangerous reactions, which in many instances cannot be completely avoided. However, adverse reactions can be minimized or even avoided if a physician is fully aware of potential risk and is prepared to appropriately handle the situation. Information on the risk of diagnostic and therapeutic procedures in allergic diseases has been accumulated in the medical literature for decades; however, except for allergen specific immunotherapy, it has never been presented in a systematic fashion. Up to now no single document addressed the risk of the most commonly used medical procedures in the allergy office nor attempted to present general requirements necessary to assure the safety of these procedures. Following review of available literature a group of allergy experts within the World Allergy Organization (WAO, representing various continents and areas of allergy expertise, presents this report on risk associated with diagnostic and therapeutic procedures in allergology and proposes a consensus on safety requirements for performing procedures in allergy offices. Optimal safety measures including appropriate location, type and required time of supervision, availability of safety equipment, access to specialized emergency services, etc. for various procedures have been recommended. This document should be useful for allergists with already established

Nuclear safety requirements for operation licensing of Egyptian research reactors

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.

2000-01-01

From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

Occupational safety and health in the Universities: fulfilling the fundamental requirement of OSHA and AELA

International Nuclear Information System (INIS)

Ismail Bahari

2000-01-01

This paper discusses the result of a survey among the universities to looks at whether such basic similarities in requirements by both Acts actually help in fulfilling and integrating the fundamental requirement of OSHA, Malaysian Occupational Safety and Health Act and AELA, Malaysian Atomic Energy Licensing Act especially through self-regulation

Identifying environmental safety and health requirements for the Fernald Environmental Restoration Management Corporation

International Nuclear Information System (INIS)

Beckman, W.H.; Cossel, S.C.; Alhadeff, N.; Lindamood, S.B.; Beers, J.A.

1994-01-01

This presentation will describe the Fernald Environmental Restoration Management Corporation's (FERMCO) Standards/Requirements Identification Documents (S/RlDs) Program, the unique process used to implement it, and the status of the program. We will also discuss the lessons learned as the program was implemented. The Department of Energy (DOE) established the Fernald site to produce uranium metals for the nation's defense programs in 1953. In 1989, DOE suspended production and, in 1991, the mission of the site was formally changed to one of environmental cleanup and restoration. The site was renamed the Fernald Environmental Management Project (FEMP). FERMCO's mission is to provide safe, early, and least-cost final clean-up of the site in compliance with all regulations and commitments. DOE has managed nuclear facilities primarily through its oversight of Management and Operating contractors. Comprehensive nuclear industry standards were absent when most DOE sites were first established, Management and Operating contractors had to apply existing non-nuclear industry standards and, in many cases, formulate new technical standards. Because it was satisfied with the operation of its facilities, DOE did not incorporate modern practices and standards as they became available. In March 1990, the Defense Nuclear Facilities Safety Board issued Recommendation 90-2, which called for DOE to identify relevant standards and requirements, conduct adequacy assessments of requirements in protecting environmental, public, and worker health and safety, and determine the extent to which the requirements are being implemented. The Environmental Restoration and Waste Management Office of DOE embraced the recommendation for facilities under its control. Strict accountability requirements made it essential that FERMCO and DOE clearly identify applicable requirements necessary, determine the requirements' adequacy, and assess FERMCO's level of compliance

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

International Nuclear Information System (INIS)

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

2005-01-01

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

49 CFR 173.473 - Requirements for foreign-made packages.

Science.gov (United States)

2010-10-01

...-3650, or by electronic mail (e-mail) to “[email protected]” If the offeror is requesting the... Transport of Radioactive Material, No. TS-R-1, ” (IBR, see § 171.7 of this subchapter) shall also comply...) 366-3753 or (202) 366-3650, or by electronic mail to “[email protected]” Each request is considered in...

Supervision of nuclear safety - IAEA requirements, accepted solutions, trends

International Nuclear Information System (INIS)

Jurkowski, M.

2007-01-01

Ten principles of the nuclear safety, based on the IAEA's standards are presented. Convention on Nuclear Safety recommends for nuclear safety landscape, the control transparency, culture safety, legal framework and knowledge preservation. Examples of solutions accepted in France, Finland, and Czech Republic are discussed. New trends in safety fundamentals and Integration Regulatory Review are presented

A study to develop the domestic functional requirements of the specific safety systems of CANDU

Energy Technology Data Exchange (ETDEWEB)

Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

2003-03-15

The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

A study to develop the domestic functional requirements of the specific safety systems of CANDU

Energy Technology Data Exchange (ETDEWEB)

Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

2001-03-15

The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

International Nuclear Information System (INIS)

2003-08-01

Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

Construction safety program for the National Ignition Facility Appendix A: Safety Requirements

International Nuclear Information System (INIS)

Cerruti, S.J.

1997-01-01

These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and construction contractors/subcontractors. The General Safety and Health rules shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S ampersand H A-1 that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Safety Rules

Construction safety program for the National Ignition Facility Appendix A: Safety Requirements

Energy Technology Data Exchange (ETDEWEB)

Cerruti, S.J.

1997-01-14

These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and construction contractors/subcontractors. The General Safety and Health rules shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S & H A-1 that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Safety Rules.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Safety requirements laid down in the Atomic Energy Law and in the Law on Immission Control

International Nuclear Information System (INIS)

Hansmann, K.

1981-01-01

The paper deals with safety requirements relating to installations, laid down in the Atomic Energy Law and in the Law on Immission Control. Actually it is a matter of how the safety requirements of sect. 7 of the Atomic Energy Law can be compared with those laid down in the sections 5 and 6 of the Federal Act for the Protection Against Nuisances. In the process, three comparative levels are examined: 1. The normative conditions concerning the licencability of hazardous installations, 2. those demands that go way beyond that in order to reduce residual risks, and 3. the licensing authorities' scope of discretion. (orig./HP) [de

Safety of High Speed Magnetic Levitation Transportation Systems - Comparison of U.S. and Foreign Safety Requirements for Application to U.S. Maglev Systems

Science.gov (United States)

1993-09-01

This report presents the results of a systematic review of the safety requirements selected for the German Transrapid : electromagnetic (EMS) type maglev system to determine their applicability and completeness with respect to the : construction and ...

Requirements and analysis of electromagnetic compatibility of safety-related instrumentation and control system in nuclear power plants

International Nuclear Information System (INIS)

Liu Sujuan

2002-01-01

The state-of-the-art instrumentation and control system and the influence of their application to the electromagnetic compatibility is analyzed. Based on the present situation of nuclear safety in China and relevant experiences from other countries, the author tries to probe into the requirements and test methods about how safety-related instrument and control system to accommodate electromagnetic interference, radio-frequency interference and power surges in the environments of nuclear power plant so as to develop Chinese safety standards

Qualification testing facility for type A, B and C packages to be used for transport and storage of radioactive materials

International Nuclear Information System (INIS)

Vieru, G.; Nistor, V.; Vasile, A.; Cojocaru, V.

2009-01-01

In accordance with the Economic Commission for Europe-Committee on inland transport (ADR- European Agreement-concerning the international carriage of dangerous goods by road, 2007 Edition) the Safety and Security of the dangerous goods class No. 7 - Radioactive Materials during transport in all different modes - by road, by rail, by sea, by inland rivers or by air - have to be ensured at a very high level. The radioactive materials (RAM) packaging have to comply to all transport conditions, routine or in accident conditions, possibly to occur during transportation operations. It is well known that the safety in the transport of RAM is dependent on packaging appropriate for the contents being shipped rather than on operational and/or administrative actions required for the package. The quality of these packages - type A, B or C has to be proved by performing qualification tests in accordance with the Romanian nuclear regulation conditions provided by CNCAN Order no. 357/22.12.2005- N orms for a Safe Transport of Radioactive Material , the IAEA Vienna Recommendation (1, 2) stipulated in the Safety standard TS-R-1- Regulation for the Safe Transport of Radioactive Material, 2005 Edition, and other applicable international recommendations. The paper will describe the components of the designed testing facilities, and the qualification testing to be performed for all type A, B and C packages subjected to the testing Quality assurance and quality controls measures taken in order to meet technical specification provided by the design are also presented and commented. The paper concludes that the new Romanian Testing Facilities for RAM packages will comply with the national safe standards as well as with the IAEA applicable recommendation provided by the TS-R-1 safety standard. (authors)

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

Science.gov (United States)

2010-07-01

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

46 CFR 53.05-1 - Safety valve requirements for steam boilers (modifies HG-400 and HG-401).

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Safety valve requirements for steam boilers (modifies HG-400 and HG-401). 53.05-1 Section 53.05-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY... requirements for steam boilers (modifies HG-400 and HG-401). (a) The pressure relief valve requirements and the...

IAEA safety standards and approach to safety of advanced reactors

International Nuclear Information System (INIS)

Gasparini, M.

2004-01-01

The paper presents an overview of the IAEA safety standards including their overall structure and purpose. A detailed presentation is devoted to the general approach to safety that is embodied in the current safety requirements for the design of nuclear power plants. A safety approach is proposed for the future. This approach can be used as reference for a safe design, for safety assessment and for the preparation of the safety requirements. The method proposes an integration of deterministic and risk informed concepts in the general frame of a generalized concept of safety goals and defence in depth. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor including small and medium sized reactors with innovative safety features.(author)

Safety class methodology

International Nuclear Information System (INIS)

Donner, E.B.; Low, J.M.; Lux, C.R.

1992-01-01

DOE Order 6430.1A, General Design Criteria (GDC), requires that DOE facilities be evaluated with respect to ''safety class items.'' Although the GDC defines safety class items, it does not provide a methodology for selecting safety class items. The methodology described in this paper was developed to assure that Safety Class Items at the Savannah River Site (SRS) are selected in a consistent and technically defensible manner. Safety class items are those in the highest of four categories determined to be of special importance to nuclear safety and, merit appropriately higher-quality design, fabrication, and industrial test standards and codes. The identification of safety class items is approached using a cascading strategy that begins at the 'safety function' level (i.e., a cooling function, ventilation function, etc.) and proceeds down to the system, component, or structure level. Thus, the items that are required to support a safety function are SCls. The basic steps in this procedure apply to the determination of SCls for both new project activities, and for operating facilities. The GDC lists six characteristics of SCls to be considered as a starting point for safety item classification. They are as follows: 1. Those items whose failure would produce exposure consequences that would exceed the guidelines in Section 1300-1.4, ''Guidance on Limiting Exposure of the Public,'' at the site boundary or nearest point of public access 2. Those items required to maintain operating parameters within the safety limits specified in the Operational Safety Requirements during normal operations and anticipated operational occurrences. 3. Those items required for nuclear criticality safety. 4. Those items required to monitor the release of radioactive material to the environment during and after a Design Basis Accident. Those items required to achieve, and maintain the facility in a safe shutdown condition 6. Those items that control Safety Class Item listed above

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

International Nuclear Information System (INIS)

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

2001-11-01

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

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

Science.gov (United States)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

2016-02-01

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

Changes in the regulations for the safe transport of radioactive material

International Nuclear Information System (INIS)

Lopez Vietri, Jorge R.; Vidal, Dora N.; Piumetti, Elsa H.; Capadona, Nancy M.

2000-01-01

The objective of this paper is to describe and to analyze the relevant changes, dealing with the design, operation and administrative requirements, to be introduced in the Revision 1 of the AR 10.16.1 standard 'Transport of radioactive material' that will be put into force on July, 1st 2001 by the Nuclear Regulatory Authority (competent authority of Argentina). In that way, the Revision 1 of the mentioned standard will be coincident with the 1996 edition (revised) of the 'Regulations for the safe transport of radioactive material', Safety Standards Series No. TS-R-1 (ST-1, revised) issued by the International Atomic Energy Agency (IAEA). (author)

49 CFR 214.507 - Required safety equipment for new on-track roadway maintenance machines.

Science.gov (United States)

2010-10-01

... maintenance machines. 214.507 Section 214.507 Transportation Other Regulations Relating to Transportation... Roadway Maintenance Machines and Hi-Rail Vehicles § 214.507 Required safety equipment for new on-track roadway maintenance machines. (a) Each new on-track roadway maintenance machine shall be equipped with: (1...

A study on the primary requirement for the safety of the Wolsong tritium removal facility

International Nuclear Information System (INIS)

Hwang, K. H.; Lee, K. J.; Jeong, C. W.

2001-01-01

Owing to the using a heavy water as a moderator and a coolant in Heavy water reactor, A large mount of tritium is produced due to a reaction of deuterium with neutron in the reactor and some of tritium is released to the environment. In Wolsong, 4 units (CANDU-600 type) Heavy water reactor is in operation. And the generated amount of tritium is increased with the increase of operational year of the Wolsong nuclear reactor. Decommissioning of the Wolsong unit 1 is expected to start at 2013. Before 2013, to reduce the workers internal radiation doses and environmental release of tritium, Tritium Removal Facility (TRF) is required and should be operated. Wolsong TRF (WTRF) is under developing stage by Korea Electric Power Corporation(KEPCO)and scheduled to start operation about 2006. Once the facility begins operation it can be contributed to the greatly reduction of tritium release to the environment and worker's expose. In this situation, study about the safety assessment method and regulatory requirement is essential for safety insurance of WTRF. And this helps the safety acquirement, successful operation and reliance of WTRF

Preparation of safety regulatory requirements for new technology like digital system

International Nuclear Information System (INIS)

Ito, Juichiro; Takita, Masami

2011-01-01

The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG (Digital Instrumentation and Control Working Group) of MDEP (Multinational Design Evaluation Program). MDEP DICWG held in OECD/NEA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency) gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance regarding digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed upon process and framework. Five general common positions are under discussion in this fiscal year. Simplicity in Design, Software Common Cause Failures, Software Tools, Data communication, Verification and Validation throughout the life cycle of safety systems using digital computers. In addition, the technical evaluation of standards of the Japan Electric Association about digital system for safety was made to support NISA (Nuclear and Industrial Safety Agency). (author)

Firefighter safety for PV systems: Overview of future requirements and protection systems

DEFF Research Database (Denmark)

Spataru, Sergiu; Sera, Dezso; Blaabjerg, Frede

2013-01-01

for operators during maintenance or fire-fighting. One of the solutions is individual module shutdown by short-circuiting or disconnecting each PV module from the PV string. However, currently no standards have been adopted either for implementing or testing these methods, or doing an evaluation of the module...... shutdown procedures. This paper gives an overview on the most recent fire - and firefighter safety requirements for PV systems, with focus on system and module shutdown systems. Several solutions are presented, analyzed and compared by considering a number of essential characteristics, including......An important and highly discussed safety issue for photovoltaic systems is that, as long as they are illuminated, a high voltage is present at the PV string terminals and cables between the string and inverters, independent of the state of the inverter's dc disconnection switch, which poses a risk...

Development of the switch requirements and architecture of a safety data communication system

International Nuclear Information System (INIS)

Jeong, K.I.; Lee, J.K.; Park, H.Y.; Koo, I.S.

2004-12-01

In accordance with digitalising the Instrumentation and Control(I and C) systems in the integral reactor, a communication network is required for effective information exchanges between the different equipment, an enhancement of the design flexibility, a simple installation and cost reduction. Generally, a communication network consists of a topology, the protocol, a communication medium, an interconnection device, etc. In this report, the development methods of switch and the architecture of a Safety Data Communication System(SDCS) are investigated and analyzed. In this report, the design requirements for switch are presented, which are the essential requirements to develop the switch in a SDCS of the SMART-P. To establish these requirements, the evaluation and analysis of the design and implementation method of the COTS switches, the architecture of SDCS and the design requirements of a SDCS were performed. At the detail design stage, these requirements will be used for the top-tier requirements, especially the design target and design basis. To develop the detail design requirements in the future, more quantitative and qualitative analyses are required. In the case of selecting the COTS switch and developing the switch, these requirements will also be used for the evaluation guide

Development of the switch requirements and architecture of a safety data communication system

Energy Technology Data Exchange (ETDEWEB)

Jeong, K.I.; Lee, J.K.; Park, H.Y.; Koo, I.S

2004-12-01

In accordance with digitalising the Instrumentation and Control(I and C) systems in the integral reactor, a communication network is required for effective information exchanges between the different equipment, an enhancement of the design flexibility, a simple installation and cost reduction. Generally, a communication network consists of a topology, the protocol, a communication medium, an interconnection device, etc. In this report, the development methods of switch and the architecture of a Safety Data Communication System(SDCS) are investigated and analyzed. In this report, the design requirements for switch are presented, which are the essential requirements to develop the switch in a SDCS of the SMART-P. To establish these requirements, the evaluation and analysis of the design and implementation method of the COTS switches, the architecture of SDCS and the design requirements of a SDCS were performed. At the detail design stage, these requirements will be used for the top-tier requirements, especially the design target and design basis. To develop the detail design requirements in the future, more quantitative and qualitative analyses are required. In the case of selecting the COTS switch and developing the switch, these requirements will also be used for the evaluation guide.

Infrastructural requirements for local implementation of safety policies: the discordance between top-down and bottom-up systems of action.

Science.gov (United States)

Timpka, Toomas; Nordqvist, Cecilia; Lindqvist, Kent

2009-03-09

Safety promotion is planned and practised not only by public health organizations, but also by other welfare state agencies, private companies and non-governmental organizations. The term 'infrastructure' originally denoted the underlying resources needed for warfare, e.g. roads, industries, and an industrial workforce. Today, 'infrastructure' refers to the physical elements, organizations and people needed to run projects in different societal arenas. The aim of this study was to examine associations between infrastructure and local implementation of safety policies in injury prevention and safety promotion programs. Qualitative data on municipalities in Sweden designated as Safe Communities were collected from focus group interviews with municipal politicians and administrators, as well as from policy documents, and materials published on the Internet. Actor network theory was used to identify weaknesses in the present infrastructure and determine strategies that can be used to resolve these. The weakness identification analysis revealed that the factual infrastructure available for effectuating national strategies varied between safety areas and approaches, basically reflecting differences between bureaucratic and network-based organizational models. At the local level, a contradiction between safety promotion and the existence of quasi-markets for local public service providers was found to predispose for a poor local infrastructure diminishing the interest in integrated inter-agency activities. The weakness resolution analysis showed that development of an adequate infrastructure for safety promotion would require adjustment of the legal framework regulating injury data exchange, and would also require rational financial models for multi-party investments in local infrastructures. We found that the "silo" structure of government organization and assignment of resources was a barrier to collaborative action for safety at a community level. It may therefore be

Comparison of Survival and Safety Requirements in European Union for Recreational Craft Inspections. A Spanish Case Study

Directory of Open Access Journals (Sweden)

J. Torralbo

2014-03-01

Full Text Available Statistical data shows that a large number of maritime accidents are related to recreational craft. For instance, in Spain, more than fifty percent of the emergencies are related to pleasure boats at sea. Recreational craft marketed in the EU must comply with harmonized technical safety and environmental requirements defined by Directive 94/25/EC, as amended in 2003. On 28 December 2013, the new recreational craft directive 2013/53/EU was published in the Official Journal of the European Union. EU Member States have until 18 January 2016 to amend their national legislation and transpose the new directive. The current directive 94/25/EC as amended by directive 2003/44/EC will be repealed on 18 January 2016, after the full application of the new text. Although this directive, there is not a clear coordination and equivalence among the EU countries according to the survival and safety equipment compulsory for recreational crafts. The main purpose of this paper is to analyze and compare the types of survey / inspections to be carried in pleasure craft (non-commercial use, periodicity and required safety equipment in some member states of the European Union. A case study of Spain is presented. From the results obtained, we can make clear that in the European Union there is a lack of coordination in this area and indicate the need to unify a common pattern in inspections and survival and safety requirements of recreational boats in the EU.

Safety standards for near surface disposal and the safety case and supporting safety assessment for demonstrating compliance with the standards

International Nuclear Information System (INIS)

Metcalf, P.

2003-01-01

The report presents the safety standards for near surface disposal (ICRP guidance and IAEA standards) and the safety case and supporting safety assessment for demonstrating compliance with the standards. Special attention is paid to the recommendations for disposal of long-lived solid radioactive waste. The requirements are based on the principle for the same level of protection of future individuals as for the current generation. Two types of exposure are considered: human intrusion and natural processes and protection measures are discussed. Safety requirements for near surface disposal are discussed including requirements for protection of human health and environment, requirements or safety assessments, waste acceptance and requirements etc

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

Generic requirements specification for qualifying a commercially available PLC for safety-related applications in nuclear power plants. Final report

International Nuclear Information System (INIS)

Ostenso, A.; May, R.

1996-12-01

This is a specification for qualifying a commercially available PLC for application to safety systems in nuclear power plants. The specifications are suitable for evaluating a particular PLC product line as a platform for safety-related applications, establishing a suitable qualification test program, and confirming that the manufacturer has a quality assurance program that is adequate for safety-related applications or is sufficiently complete that, with a reasonable set of compensatory actions, it can be brought into conformance. The specification includes requirements for: (1) quality assurance measures applied to the qualification activities, (2) documentation to support the qualification, and (3) documentation to provide the information needed for applying the qualified PLC platform to a specific application. The specifications are designed to encompass a broad range of safety applications; however, qualifying a particular platform for a different range of applications can be accomplished by appropriate adjustments to the requirements

Data requirements for the Ferrocyanide Safety Issue developed through the data quality objectives process

International Nuclear Information System (INIS)

Meacham, J.E.; Cash, R.J.; Dukelow, G.T.; Babad, H.; Buck, J.W.; Anderson, C.M.; Pulsipher, B.A.; Toth, J.J.; Turner, P.J.

1994-08-01

This document records the data quality objectives (DQO) process applied to the Ferrocyanide Safety Issue at the Hanford Site. Specifically, the major recommendations and findings from this Ferrocyanide DQO process are presented. The decision logic diagrams and decision error tolerances also are provided. The document includes the DQO sample-size formulas for determining specific tank sampling requirements, and many of the justifications for decision thresholds and decision error tolerances are briefly described. More detailed descriptions are presented in other Ferrocyanide Safety Program companion documents referenced in this report. This is a living document, and the assumptions contained within will be refined as more data from sampling and characterization become available

Evaluation of Activity Concentration Values and Doses due to the Transport of Low Level Radioactive Material

Energy Technology Data Exchange (ETDEWEB)

Rawl, Richard R [ORNL; Scofield, Patricia A [ORNL; Leggett, Richard Wayne [ORNL; Eckerman, Keith F [ORNL

2010-04-01

The International Atomic Energy Agency (IAEA) initiated an international Coordinated Research Project (CRP) to evaluate the safety of transport of naturally occurring radioactive material (NORM). This report presents the United States contribution to that IAEA research program. The focus of this report is on the analysis of the potential doses resulting from the transport of low level radioactive material. Specific areas of research included: (1) an examination of the technical approach used in the derivation of exempt activity concentration values and a comparison of the doses associated with the transport of materials included or not included in the provisions of Paragraph 107(e) of the IAEA Safety Standards, Regulations for the Safe Transport of Radioactive Material, Safety Requirements No. TS-R-1; (2) determination of the doses resulting from different treatment of progeny for exempt values versus the A{sub 1}/A{sub 2} values; and (3) evaluation of the dose justifications for the provisions applicable to exempt materials and low specific activity materials (LSA-I). It was found that the 'previous or intended use' (PIU) provision in Paragraph 107(e) is not risk informed since doses to the most highly exposed persons (e.g., truck drivers) are comparable regardless of intended use of the transported material. The PIU clause can also have important economic implications for co-mined ores and products that are not intended for the fuel cycle but that have uranium extracted as part of their industrial processing. In examination of the footnotes in Table 2 of TS-R-1, which identifies the progeny included in the exempt or A1/A2 values, there is no explanation of how the progeny were selected. It is recommended that the progeny for both the exemption and A{sub 1}/A{sub 2} values should be similar regardless of application, and that the same physical information should be used in deriving the limits. Based on the evaluation of doses due to the transport of low

Multimegawatt Space Reactor Safety

International Nuclear Information System (INIS)

Stanley, M.L.

1989-01-01

The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

Preparation of safety regulatory requirements for new technology like digital system

International Nuclear Information System (INIS)

2012-01-01

The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG of MDEP. MDEP DICWG held in OECD/NEA gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance on digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed process and framework. The following two general common positions are discussed and to be issued in this fiscal year. Verification and Validation throughout the life cycle of safety systems using digital computers. The Impact of Cyber Security Features on Digital I and C Safety Systems. (author)

Population and colony-level determinants of tertiary sex ratio in the declining barn swallow.

Directory of Open Access Journals (Sweden)

Nicola Saino

Full Text Available Sex ratio of adults (tertiary sex ratio, TSR is a major feature of animal populations with consequences for their behaviour, genetic structure and viability. Spatial and temporal variation in TSR occurs within species but the mechanisms behind it are poorly understood. In this long-term study of a declining population of a socially monogamous, colonial, migratory bird, the barn swallow (Hirundo rustica, we first analyzed population-level variation in TSR ( =  proportion of males of yearlings at sexual maturation in relation to ecological conditions as gauged by annual survival rate of adults. TSR was male-biased both among yearlings and older individuals, but male bias of yearlings was more pronounced after years with larger decline in adult survival. Thus, male offspring were less susceptible to the adverse ecological conditions that cause increased mortality. Dispersal and settling site decisions can have major consequences on fitness via the effects of local TSR on mating and sperm competition. Breeding barn swallows are highly philopatric while natal dispersal is high and, together with mortality, is the main determinant of colony TSR. We thus also investigated the mechanisms of breeding colony choice by yearlings and found that TSR of new-settlers in a given colony and year was negatively predicted by TSR of returning, early arriving older individuals in that year, but not by overall TSR at the colony in the previous year. This suggests that in our male-biased population new-settler males respond to local TSR upon arrival to choose the sites with larger breeding opportunities. Hence, variation in ecological conditions as reflected by adult survival can shift the TSR of individuals recruiting into a local population, with potentially various demographic consequences. However, breeding site choice based on TSR tends to homogenize TSR at a population level likely by facilitating settling of dispersing males in colonies with less male

Lithuanian requirements for ageing management of systems and components important to safety of nuclear power plant

International Nuclear Information System (INIS)

Ramanauskiene, A.

2000-01-01

In this paper the Lithuanian requirements for ageing management of systems and components important to safety of Ignalina nuclear power plant (two RBMK-1500 water-cooled graphite moderated channel-type power reactors) are presented

Combined evaluation. Plutonium transports in France. Problems of safety and reliability of transport container FS47; Evaluation Conjointe. Transports de plutonium en France. Problemes de surete et de securite du container de transport FS47

Energy Technology Data Exchange (ETDEWEB)

Marignac, Y.; Coeytaux, X. [Wise-Paris, 75 (France); Large, J.H. [Nuclear Engineer, Large and Associates, Londres (United Kingdom)

2004-09-15

This report concerns the safety and the protection of plutonium dioxide transported from Cogema La Hague to the mixed oxide fuel plant of Marcoule and Cadarache. The French approach of the transport safety is based on the combining of two essential principles: the first one affirms that the performances of the FS47 container in regard of containment (norms TS-R-1 from IAEA for the accidental conditions) is conceived to resist in any situation even terrorism or sabotage. In fact, the IAEA norm follows a probabilistic study without a voluntary attack such a terrorist one. The second principle rests on the ability to prevent the treat of terrorism acts, because of a secrecy policy on the plutonium transport. It appeared that the Green peace association has succeeded several times to know exactly the hours, the trips of the plutonium transport and this simple thing raises more questions than it solves. (N.C.)

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

25 Years of Community Activities towards Harmonization of Nuclear Safety Criteria and Requirements - Achievements and Prospects

International Nuclear Information System (INIS)

Lillington, J.N.; Turland, B.D.; Haste, T.J.; Seiler, J.M.; Carretero, A.; Perez, T.; Geutges, A.; Van Hienen, J.F.A.; Jehee, J.N.T.; Sehgal, B.R.; Mattila, L.; Holmstrom, H.; Karwat, H.; Maroti, L.; Toth, I.; Husarcek, J.

2001-10-01

The main objective was to advise the EC on future challenges and opportunities in terms of enhanced co-operation in the area of nuclear safety and harmonization of safety requirements and practices in an enlarged European Union. The activities were divided into 3 sub-tasks as follows: part A, to prepare an analysis, synthesis and assessment of the main achievements from Community activities related to the Resolutions on the technological problems of nuclear safety of 1975 and 1992, with due consideration for related research activities; part B, to prepare an overview of safety philosophies and practices in EU Member States, taking account of their specific national practices in terms of legal framework, type and age of operating nuclear reactors; part C, to provide elements of a strategy for future activities in the frame of the Council Resolutions, with particular attention to the context of enlargement of the EU. (author)

Development of the environmental qualification safety requirement matrix for the containment system of in-service CANDU reactors

International Nuclear Information System (INIS)

Chun, R.M.; Low, J.; Sobolewski, J.

1994-01-01

Over the last several years, Ontario Hydro Nuclear (OHN) has placed increasing emphasis on environmental qualification (EQ) at its Pickering and Bruce NGS A and B nuclear generating stations (NGSs). The program currently underway (at the time of the conference) builds upon the experience gained from the extensive Darlington NGS EQ experience and from EQ programs conducted by other utilities. Some of the major steps of the OHN EQ program include: defining Safety Requirement Matrices (SRMs), establishing environmental conditions, developing an EQ List, conducting an EQ Assessment and maintaining Operational EQ Assurance during the plant life. The SRM identifies safety related components, their required safety functions and their mission times for each postulated design basis accident (DBA). This is a critical step, as the SRM defines the equipment that requires assurance of EQ and precise requirements must be provided to ensure a cost effective EQ program. This paper describes the development of the SRMs for the containment system of the Bruce stations. The introductory section briefly discusses how the industry has dealt with equipment qualification as it has evolved and the role of the SRMs in the OHN EQ Program. In Section 2, the preparation of the SRM is described along with the applicable ground rules used. The results of the application of the SRM preparation guidelines to the containment system are discussed in Section 3. A summary of the major findings and conclusions is presented. 3 refs., 3 figs

Regulatory Oversight of Safety Culture in Finland: A Systemic Approach to Safety

International Nuclear Information System (INIS)

Oedewald, P.; Väisäsvaara, J.

2016-01-01

In Finland the Radiation and Nuclear Safety Authority STUK specifies detailed regulatory requirements for good safety culture. Both the requirements and the practical safety culture oversight activities reflect a systemic approach to safety: the interconnections between the technical, human and organizational factors receive special attention. The conference paper aims to show how the oversight of safety culture can be integrated into everyday oversight activities. The paper also emphasises that the scope of the safety culture oversight is not specific safety culture activities of the licencees, but rather the overall functioning of the licence holder or the new build project organization from safety point of view. The regulatory approach towards human and organizational factors and safety culture has evolved throughout the years of nuclear energy production in Finland. Especially the recent new build projects have highlighted the need to systematically pay attention to the non-technical aspects of safety as it has become obvious how the HOF issues can affect the design processes and quality of construction work. Current regulatory guides include a set of safety culture related requirements. The requirements are binding to the licence holders and they set both generic and specific demands on the licencee to understand, monitor and to develop safety culture of their own organization but also that of their supplier network. The requirements set for the licence holders has facilitated the need to develop the regulator’s safety culture oversight practices towards a proactive and systemic approach.

Describing and analyzing effects of international differences in food safety requirements -the case of the EU versus US-

NARCIS (Netherlands)

Bremmers, H.J.; Meulen, van der B.M.J.; Poppe, K.J.; Wijnands, J.H.M.

2010-01-01

Abstract This paper compares SPS-requirements of the USA and of the EU from the perspective of the processing establishment, and analyzes the consequences of differences for national as well as firm policies. Differences in safety requirements may impede the competitiveness of the food industry.

Development of U.S. Government General Technical Requirements for UAS Flight Safety Systems Utilizing the Iridium Satellite Constellation

Science.gov (United States)

Murray, Jennifer; Birr, Richard

2010-01-01

This slide presentation reviews the development of technical requirements for Unmanned Aircraft Systems (UAS) utilization of the Iridium Satellite Constellation to provide flight safety. The Federal Aviation Authority (FAA) required an over-the-horizon communication standard to guarantee flight safety before permitting widespread UAS flights in the National Air Space (NAS). This is important to ensure reliable control of UASs during loss-link and over-the-horizon scenarios. The core requirement was to utilize a satellite system to send GPS tracking data and other telemetry from a flight vehicle down to the ground. Iridium was chosen as the system because it is one of the only true satellite systems that has world wide coverage, and the service has a highly reliable link margin. The Iridium system, the flight modems, and the test flight are described.

Infrastructural requirements for local implementation of safety policies: the discordance between top-down and bottom-up systems of action

Directory of Open Access Journals (Sweden)

Lindqvist Kent

2009-03-01

Full Text Available Abstract Background Safety promotion is planned and practised not only by public health organizations, but also by other welfare state agencies, private companies and non-governmental organizations. The term 'infrastructure' originally denoted the underlying resources needed for warfare, e.g. roads, industries, and an industrial workforce. Today, 'infrastructure' refers to the physical elements, organizations and people needed to run projects in different societal arenas. The aim of this study was to examine associations between infrastructure and local implementation of safety policies in injury prevention and safety promotion programs. Methods Qualitative data on municipalities in Sweden designated as Safe Communities were collected from focus group interviews with municipal politicians and administrators, as well as from policy documents, and materials published on the Internet. Actor network theory was used to identify weaknesses in the present infrastructure and determine strategies that can be used to resolve these. Results The weakness identification analysis revealed that the factual infrastructure available for effectuating national strategies varied between safety areas and approaches, basically reflecting differences between bureaucratic and network-based organizational models. At the local level, a contradiction between safety promotion and the existence of quasi-markets for local public service providers was found to predispose for a poor local infrastructure diminishing the interest in integrated inter-agency activities. The weakness resolution analysis showed that development of an adequate infrastructure for safety promotion would require adjustment of the legal framework regulating injury data exchange, and would also require rational financial models for multi-party investments in local infrastructures. Conclusion We found that the "silo" structure of government organization and assignment of resources was a barrier to

Development of Occupational Safety and Health Requirement Management System (OSHREMS Software Using Adobe Dreamweaver CS5 for Building Construction Project

Directory of Open Access Journals (Sweden)

Abas Nor Haslinda

2017-01-01

Full Text Available The construction industry sector is considered as being risky with frequent and high accident rate. According to Social Security Organization (SOCSO, the construction accidents has arisen from time to time. Construction Industry Development Board (CIDB has developed the Safety and Health Assessment System in Construction (SHASSIC for evaluating the performance of a contractor in construction project by setting out the safety and health management and practices, however the requirement checklist provided is not comprehensive. Therefore, this study aims to develop a software system for facilitating OSH in building construction project, namely OSH requirements management system (OSHREMS, using Adobe Dreamweaver CS5 and Sublime Text as PHP editor. The results from a preliminary study which was conducted through interviews showed that, the respondents were only implementing the basic requirements that comply with legislations, with the absence of appropriate and specific guideline in ensuring occupational safety and health (OSH at the workplace. The tool will be benefits for contractors and other parties to effectively manage the OSH requirements for their projects based on project details.

Structural Design Requirements and Factors of Safety for Spaceflight Hardware: For Human Spaceflight. Revision A

Science.gov (United States)

Bernstein, Karen S.; Kujala, Rod; Fogt, Vince; Romine, Paul

2011-01-01

This document establishes the structural requirements for human-rated spaceflight hardware including launch vehicles, spacecraft and payloads. These requirements are applicable to Government Furnished Equipment activities as well as all related contractor, subcontractor and commercial efforts. These requirements are not imposed on systems other than human-rated spacecraft, such as ground test articles, but may be tailored for use in specific cases where it is prudent to do so such as for personnel safety or when assets are at risk. The requirements in this document are focused on design rather than verification. Implementation of the requirements is expected to be described in a Structural Verification Plan (SVP), which should describe the verification of each structural item for the applicable requirements. The SVP may also document unique verifications that meet or exceed these requirements with NASA Technical Authority approval.

Safety requirement of the nuclear power plants, after TMI-2 accident and their possible implementation on Bushehr NPP

International Nuclear Information System (INIS)

Mirhabibi, N.; Tochai, M.T.M.; Ashrafi, A.; Farnoudi, E.

1985-01-01

Based on the lessons learned from the TMI-2 accident and other research and developments, many improvements have been required for the design, manufacturing and operation of nuclear power plants in recent years. These requirements have already been implemented to the plants in operation and considered as new safety requirements for new plants. In the present paper these requirements and their possible implementation on Bushehr NPP are discussed. (Author)

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (French Edition); Cadre gouvernemental, legislatif et reglementaire de la surete. Prescriptions generales de surete. Partie 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-11-15

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Spanish Edition); Marco gubernamental, juridico y regulador para la seguridad. Requisitos de Seguridad Generales. Parte 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-11-15

The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

Assessment of the impact of dipped guideways on urban rail transit systems: Ventilation and safety requirements

Science.gov (United States)

1982-01-01

The ventilation and fire safety requirements for subway tunnels with dipped profiles between stations as compared to subway tunnels with level profiles were evaluated. This evaluation is based upon computer simulations of a train fire emergency condition. Each of the tunnel configurations evaluated was developed from characteristics that are representative of modern transit systems. The results of the study indicate that: (1) The level tunnel system required about 10% more station cooling than dipped tunnel systems in order to meet design requirements; and (2) The emergency ventilation requirements are greater with dipped tunnel systems than with level tunnel systems.

Occupational exposure following Yttrium-90 microspheres SIR therapy

International Nuclear Information System (INIS)

Fontaine, P.; Boirie, G.; Dieudonne, A.; Leguludec, D.; Lebtahi, R.; Ben Reguiga, M.

2015-01-01

Full text of publication follows. Introduction: Selective Internal Radiation Therapy (SIRT) is a promising technique for solid hepatic neoplasms treatment. SIRT consists in implanting radioactive microspheres (RMS) in targeted hepatic lesions via femoral artery. Two RMS-Therapsheres [glass-microspheres/TSR] and Sir-Spheres [resin-spheres/SSR]- are marketed in the European market, both radiolabeled with Yttrium-90. The objective of this study is to assess occupational exposure for nuclear medicine, radiology and clinical staff involved in Y 90 -RMS preparation and implantation. Materials and methods: The study was conducted on 20 patients treated for Hepato-Cellular Carcinoma: 10 treated with TSR and 10 with SSR. Dose rate (DR, mSv/h) or absorbed doses (mSv) measurements were made during all steps of TSR and SSR handling: sources receipt and unpacking, preparation, transport to radiology, implantation, and patient care. Measurements were made with portable ionization chamber(Babyline/Nardeux), spectrometer(FieldSpect/Aries), digital dosimeter (NED/Unfors) and operational dosimeter (Mk2/Siemens). Values were expressed as mean±SD. Results: patients received of 1.8 GBq to 3.1 GBq of TSR and 0.55 GBq to 2.4 GBq of SSR. TSR were delivered ready-to-use with the prescribed activity. For SSR only one activity was commercially available and shipped (3 GBq at calibration-time)requiring a preparation step to adjust needed activity. DR measured during RMS was 1723 ± 157 μSv/h SSR and 1189 ± 92 μSv/h for TSR. When preparing spheres in radiopharmacy, fingers and whole body doses were respectively 8326 ± 2360 μSv and 12.3 ± 5,2 μSv for Sir-Spheres vs. 33.5 ± 7.8 μSv and 1.1 ± 0.3 μSv for TSR. DR in contact with carrying case during RMS transfer to radiology were 299 ± 102 μSv/h for SSR and 5.3 ± 1.2 μSv/h for TSR. During RMS infusion, radiologist's finger doses were limited to 3.6 ± 1 μSv for SSR and 0.7± 0.3 μSv for TSR. Finally, following RMS

Reactivity requirements and safety systems for heavy water reactors

International Nuclear Information System (INIS)

Kati, S.L.; Rustagi, R.S.

1977-01-01

The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

Preparation of safety regulatory requirements for new technology like digital system

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG of MDEP. MDEP DICWG held in OECD/NEA gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance on digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed process and framework. The following four general common positions have been discussed in this fiscal year. The Treatment of Common Cause Failure Resulting from Software within Digital Safety Systems, The Treatment of Hardware Description Language(HDL) Programmed Devices for Use in Nuclear Safety System, Factory Acceptance Test and Site Acceptance Test, The Use of Automatic Tests to Perform Surveilance for Digital Systems. (author)

VDMA contribution to functional safety of turbomachinery. Required risk reduction by safety functions for steam turbines; VDMA-Beitrag zur Funktionalen Sicherheit von Turbomaschinen. Notwendige Risikoreduktion durch Schutzfunktionen fuer Dampfturbinen

Energy Technology Data Exchange (ETDEWEB)

Wuest, Bernhard [Alstom Power Systems GmbH, Mannheim (Germany); Zelinger, Matthias [VDMA Power Systems, Frankfurt am Main (Germany); Havemann, Juergen [Siemens AG, Muelheim an der Ruhr (Germany). Energy Sector; Potten, Christian [MAN Diesel und Turbo SE, Oberhausen (Germany)

2011-07-01

Turbomachinery in power plants and industrial plants has to satisfy high safety standards. To meet these requirements, mechanical, hydraulic and electromechanical components have been used, most of them well-established already for decades. In recent years new standards for functional safety have been developed which address different target groups (IEC 61 528/511 for process industry IEC 62061 and ISO 13849 for mechanical engineering). The Working Panel 'Functional Safety of Turbomachinery' of VDMA defines rules for turbomachinery that will be presented with their background. (orig.)

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.

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

Requirement and prospect of nuclear data activities for nuclear safety

International Nuclear Information System (INIS)

Kimura, Itsuro

2000-01-01

Owing to continuous efforts by the members of JNDC (Japanese Nuclear Data Committee) and Nuclear Data Center in JAERI (Japan Atomic Energy Research Institute), several superb evaluated nuclear data files, such as JENDL, FP (fission product) yields and decay heat, have been compiled in Japan and opened to the world. However, they are seldom adopted in safety design and safety evaluation of light water reactors and are hardly found in related safety regulatory guidelines and standards except the decay heat. In this report, shown are a few examples of presently used nuclear data in the safety design and the safety evaluation of PWRs (pressurized water reactors) and so forth. And then, several procedures are recommended in order to enhance more utilization of Japanese evaluated nuclear data files for nuclear safety. (author)

Safety design guide for safety related systems for CANDU 9

International Nuclear Information System (INIS)

Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

1996-03-01

In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new

Safety design guide for safety related systems for CANDU 9

Energy Technology Data Exchange (ETDEWEB)

Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

1996-03-01

In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.

An Optimal Analysis in Wireless Powered Full-duplex Relaying Network

Directory of Open Access Journals (Sweden)

K.-T. Nguyen

2017-04-01

Full Text Available Wireless-powered cellular networks (WPCNs are currently being investigated to exploit the reliability and improve battery lifetime of mobile users. This paper investigates the energy harvesting structure of the full-duplex relaying networks. By using the time switching based relaying (TSR protocol and Amplify-and-Forward (AF model in delay-limited transmission scheme, we propose the closed-form expression of the outage probability and then calculate the optimal throughput. An important result can be taken obviously that the time fraction in TSR, the position of relay, the noise as well as the energy conversation impacting on the outage probability as well as the optimal throughput. By Monte Carlo simulation, the numerical results indicate an effective relaying strategy in full-duplex cooperative systems. Finally, we provide fundamental design guidelines for selecting time fraction in TSR that satisfies the requirements of a practical relaying system.

The tendency of medical electrical equipment - IEC 60601-2-54: Particular requirements for the basic safety and essential performance of x-ray equipment for radiography and radioscopy

International Nuclear Information System (INIS)

Roh, Young Hoon; Kim, Jung Min

2015-01-01

Medical electrical equipment - Part 1: General requirement for basic safety and essential performance of MFDS was revised as 3th edition and Medical electrical equipment Part 2-54: Particular requirements for the basic safety and essential performance of X-ray equipment will be expected to be announced as notification. Therefore this technical report was written to introduce provision of the particular requirements, replacement, addition, amendment. The purpose of this particular requirements is to secure requirements for basic safety and essential performance of X-ray equipment for radiography and radioscopy. X-ray high voltage generator, mechanical protective device, protection against radiation is included in this particular requirements. Medical electrical equipment - Part 1, Part 1-2, Part 1-3 is applied to this particular requirements. If the requirements is announced as notification, It is expected to widen understanding for basic safety and essential performance of X-ray equipment for radiography and radioscopy and play a part to internationalize of medical equipment

Special safety requirements applied to Brazilian nuclear power plant

International Nuclear Information System (INIS)

Lepecki, W.P.S.; Hamel, H.J.E.; Koenig, N.; Vieira, P.C.R.; Fritzsche, J.C.

1981-01-01

Some safety aspects of the Angra 2 and 3 nuclear power plants are presented. An analysis of the civil and mechanical project of these nuclear power plant having in view a safety analysis is done. (E.G.) [pt

76 FR 303 - Pipeline Safety: Safety of On-Shore Hazardous Liquid Pipelines

Science.gov (United States)

2011-01-04

... leak detection requirements for all pipelines; whether to require the installation of emergency flow... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part 195 [Docket ID PHMSA-2010-0229] RIN 2137-AE66 Pipeline Safety: Safety of On-Shore Hazardous Liquid...

Thermographic inspection of pipes, tanks, and containment liners

Energy Technology Data Exchange (ETDEWEB)

Renshaw, Jeremy B., E-mail: jrenshaw@epri.com; Muthu, Nathan [Electric Power Research Institute, 1300 West WT Harris Blvd., Charlotte, NC 28262 (United States); Lhota, James R.; Shepard, Steven M., E-mail: sshepard@thermalwave.com [Thermal Wave Imaging, 845 Livernois St., Ferndale, MI 48220 (United States)

2015-03-31