Identification and assessment of containment and release management strategies for a BWR Mark II containment

International Nuclear Information System (INIS)

Lin, C.C.; Lehner, J.R.

1992-06-01

Accident management strategies that have the potential to maintain containment integrity and control or mitigate the release of radioactivity following a severe accident at a boiling water reactor with a Mark 2 type of containment are identified and evaluated. The strategies are referred to as containment and release strategies. Using information available from probabilistic risk assessments and other existing severe accident research, and employing simplified containment and release event trees, this report identified the challenges a Mark 2 containment may encounter during a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenge. By means of a safety objective tree, the strategies are linked to the general safety objectives of containment and release management. As part of the assessment process, the strategies are applied to certain severe accident sequence categories deemed important to a Mark 2 containment. These sequence categories exhibit one or more of the following characteristics: high probability of core damage, high consequences, lead to a number of challenges, and involve the failure of multiple systems. The Limerick Generating Station is used as a representative Mark 2 plant to illustrate plant specifics in this report

Assessment of integrity of structures containing cracks

International Nuclear Information System (INIS)

Townley, C.H.A.

1976-01-01

The object of the investigations is to provide a method of assessing the safety and integrity of structures containing cracklike defects. Estimated load at which the cracked structure will fail is compared with the highest load likely to be applied in service

Safety and reliability assessment

International Nuclear Information System (INIS)

1979-01-01

This report contains the papers delivered at the course on safety and reliability assessment held at the CSIR Conference Centre, Scientia, Pretoria. The following topics were discussed: safety standards; licensing; biological effects of radiation; what is a PWR; safety principles in the design of a nuclear reactor; radio-release analysis; quality assurance; the staffing, organisation and training for a nuclear power plant project; event trees, fault trees and probability; Automatic Protective Systems; sources of failure-rate data; interpretation of failure data; synthesis and reliability; quantification of human error in man-machine systems; dispersion of noxious substances through the atmosphere; criticality aspects of enrichment and recovery plants; and risk and hazard analysis. Extensive examples are given as well as case studies

Safety Assessment of Talc as Used in Cosmetics.

Science.gov (United States)

Fiume, Monice M; Boyer, Ivan; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan

2015-01-01

The Cosmetic Ingredient Review Expert Panel (Panel) assessed the safety of talc for use in cosmetics. The safety of talc has been the subject of much debate through the years, partly because the relationship between talc and asbestos is commonly misunderstood. Industry specifications state that cosmetic-grade talc must contain no detectable fibrous, asbestos minerals. Therefore, the large amount of available animal and clinical data the Panel relied on in assessing the safety of talc only included those studies on talc that did not contain asbestos. The Panel concluded that talc is safe for use in cosmetics in the present practices of use and concentration (some cosmetic products are entirely composed of talc). Talc should not be applied to the skin when the epidermal barrier is missing or significantly disrupted. © The Author(s) 2015.

Safety margins associated with containment structures under dynamic loading

International Nuclear Information System (INIS)

Lu, S.C.

1978-01-01

A technical basis for assessing the true safety margins of containment structures involved with MARK I boiling water reactor reevaluation activities is presented. It is based on the results of a plane-strain, large displacement, elasto-plastic, finite-element analysis of a thin cylindrical shell subjected to external and internal pressure pulses. An analytical procedure is presented for estimating the ultimate load capacity of the thin shell structure, and subsequently, for quantifying the design margins of safety for the type of loads under consideration. For defining failure of structures, a finite strain failure criterion is derived that accounts for multiaxiality effects

Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

International Nuclear Information System (INIS)

2000-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

Simplified safety and containment systems for the iris reactor

International Nuclear Information System (INIS)

Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

2001-01-01

The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

Energy Technology Data Exchange (ETDEWEB)

Lin, Feng; Li, Hong Zhi [Dept. Structural Engineering, Tongji University, Shanghai (China)

2017-08-15

Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA) of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking) and Limit State II (concrete crushing) when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

Safety assessment of food and herbal products containing hepatotoxic pyrrolizidine alkaloids: interlaboratory consistency and the importance of N-oxide determination.

Science.gov (United States)

Cao, Yu; Colegate, Steven M; Edgar, John A

2008-01-01

Two recent mass spectrometry-based reports concerning Senecio scandens yielded remarkably dissimilar pyrrolizidine alkaloid constituents. In both studies, and in a related analysis of Senecio scandens and Tussilago farfara using micellar electrokinetic chromatography, the presence of hazardous N-oxides of the alkaloids was either not considered or was inadequately considered. This raises concerns about the effectiveness of the methodologies used in these, and similar, studies in assessing the pyrrolizidine alkaloid content and the safety of food, food supplements and medicines for human use. To highlight essential analytical requirements for confident assessment of pyrrolizidine alkaloid-related safety of food and herbal products for human use. Direct infusion-ESI MS and HPLC-ESI MS were used to analyse samples derived from liquid-liquid partitioning experiments and from strong cation exchange, solid-phase extraction of pyrrolizidine alkaloids and their N-oxides. A simple solvent partitioning experiment using pure senecionine and senecionine-N-oxide, two constituents reported in one of the mass spectrometry-based studies of S. scandens, clearly demonstrated the inadequacy of the reported method to detect and quantitate hazardous pyrrolizidine alkaloid N-oxide components. A preliminary LCMS analysis of commercially-prepared extracts of comfrey roots (Symphytum officinale and S. uplandicum s. l.) was used as a model to highlight the analytical importance of N-oxides in the safety assessment of pyrrolizidine alkaloid-containing medicinal herbs. This study highlighted significant differences in the reported identification of pyrrolizidine alkaloids from the same plant species, and clearly demonstrated the inadequacy of some procedures to include N-oxides in the assessment of pyrrolizidine alkaloid-related safety of food and herbal products.

Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

Environment, safety and health progress assessment manual

International Nuclear Information System (INIS)

1992-12-01

On June 27, 1989, the Secretary of Energy announced a 1O-Point Initiative to strengthen environment,safety, and health (ES ampersand H) programs, and waste management activities at involved conducting DOE production, research, and testing facilities. One of the points independent Tiger Team Assessments of DOE operating facilities. The Office of Special Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This volume contains appendices to the Environment, Safety and Health Progress Assessment Manual

Safety analysis of nuclear containment vessels subjected to strong earthquakes and subsequent tsunamis

Directory of Open Access Journals (Sweden)

Feng Lin

2017-08-01

Full Text Available Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking and Limit State II (concrete crushing when the PGAs were in a range of 0.8–1.1g and 1.2–1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

Initial performance assessment of the Westinghouse AP600 containment design and related safety issues

International Nuclear Information System (INIS)

Nicolette, V.F.; Washington, K.E.; Tills, J.L.

1991-01-01

This work summarizes the Westinghouse AP600 advanced reactor design assessment calculations performed to date with the CONTAIN code. Correlations for modeling the important heat transfer phenomena are discussed as well. A CONTAIN model of the AP600 was constructed for design basis accident (DBA) calculations. Insights gained from modeling of the smaller-scale Westinghouse Integral Test Facility were incorporated in the development of the AP600 model. The results of the DBA calculations are compared to the results of other researchers to serve as a point of reference for future severe accident calculations. The CONTAIN calculations are reviewed to examine several parameters/phenomena of interest. The results of the calculations are also used to identify limitations of the CONTAIN code regarding application to advanced reactor containment designs. The most recent heat transfer correlations available in the literature are assessed for use in the flow regimes and geometries applicable to the AP600. Use of one of these correlations in CONTAIN may allow for a more accurate assessment of the AP600

Seismic reliability assessment methodology for CANDU concrete containment structures

International Nuclear Information System (INIS)

Stephens, M.J.; Nessim, M.A.; Hong, H.P.

1995-05-01

A study was undertaken to develop a reliability-based methodology for the assessment of existing CANDU concrete containment structures with respect to seismic loading. The focus of the study was on defining appropriate specified values and partial safety factors for earthquake loading and resistance parameters. Key issues addressed in the work were the identification of an approach to select design earthquake spectra that satisfy consistent safety levels, and the use of structure-specific data in the evaluation of structural resistance. (author). 23 refs., 9 tabs., 15 figs

Safety design guides for containment extension 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 containment extension describes the containment isolation philosophy and containment extension requirements. The metal extensions and components falling within the scope of ASME Section III are classified in accordance with the CAN/CSA-N285.0 and CAN/CSA-N285.3. The special consideration for the leak monitoring capability, seismic qualification and inspection requirements for containment extensions, etc., are defined in this design guide. In addition, the containment isolation systems are defined and summarized schematically in appendix A. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. (Author) .new

Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume I

International Nuclear Information System (INIS)

2000-12-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume II

International Nuclear Information System (INIS)

2000-12-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

International Nuclear Information System (INIS)

Jordanov, M.

1999-01-01

The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

Structural Integrity Assessment of Reactor Containment Subjected to Aircraft Crash

International Nuclear Information System (INIS)

Kim, Junyong; Chang, Yoonsuk

2013-01-01

When an accident occurs at the NPP, containment building which acts as the last barrier should be assessed and analyzed structural integrity by internal loading or external loading. On many occasions that can occur in the containment internal such as LOCA(Loss Of Coolant Accident) are already reflected to design. Likewise, there are several kinds of accidents that may occur from the outside of containment such as earthquakes, hurricanes and strong wind. However, aircraft crash that at outside of containment is not reflected yet in domestic because NPP sites have been selected based on the probabilistic method. After intentional aircraft crash such as World Trade Center and Pentagon accident in US, social awareness for safety of infrastructure like NPP was raised world widely and it is time for assessment of aircraft crash in domestic. The object of this paper is assessment of reactor containment subjected to aircraft crash by FEM(Finite Element Method). In this paper, assessment of structural integrity of containment building subjected to certain aircraft crash was carried out. Verification of structure integrity of containment by intentional severe accident. Maximum stress 61.21MPa of horizontal shell crash does not penetrate containment. Research for more realistic results needed by steel reinforced concrete model

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

International Nuclear Information System (INIS)

Staples, A.T.

1992-06-01

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

Safety and efficacy of personal care products containing colloidal oatmeal.

Science.gov (United States)

Criquet, Maryline; Roure, Romain; Dayan, Liliane; Nollent, Virginie; Bertin, Christiane

2012-01-01

Colloidal oatmeal is a natural ingredient used in the formulation of a range of personal care products for relief of skin dryness and itchiness. It is also used as an adjunctive product in atopic dermatitis. The safety of personal care products used on vulnerable skin is of particular importance and the risk of developing further skin irritations and/or allergies should be minimized. In a series of studies, we tested the safety of personal care products containing oatmeal (creams, cleansers, lotions) by assessing their irritant/allergenic potential on repeat insult patch testing, in safety-in-use and ocular studies using subjects with nonsensitive and sensitive skin. We also tested the skin moisturizing and repair properties of an oatmeal-containing skin care product for dry skin. We found that oatmeal-containing personal care products had very low irritant potential as well as a very low allergenic sensitization potential. Low-level reactions were documented in 1.0% of subjects during the induction phase of repeat insult patch testing; one of 2291 subjects developed a persistent but doubtful low-level reaction involving edema during the challenge phase in repeat insult patch testing. No allergies were reported by 80 subjects after patch testing after in-use application. Sustained skin moisturizing was documented in subjects with dry skin that lasted up to 2 weeks after product discontinuation. Our results demonstrate that colloidal oatmeal is a safe and effective ingredient in personal care products. No allergies were reported by consumers of 445,820 products sold during a 3-year period.

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

The Nirex safety assessment research programme: annual report for 1986/87

International Nuclear Information System (INIS)

Cooper, M.J.; Hodgkinson, D.P.

1987-05-01

This report describes research relating to the underground disposal of low-level and intermediate-level radioactive wastes, to provide information for post-emplacement radiological safety assessment. Topics reported are solubility and sorption, organic degradation, microbial activity, leaching, the corrosion of containers, and radionuclide migration studies. Properties of clays, slates, colloids and uranium disequilibrium are studied. Mathematical modelling to support the safety assessment of radioactive waste disposal is also studied. (U.K.)

HSE's safety assessment principles for criticality safety

International Nuclear Information System (INIS)

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

2008-01-01

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

Safety assessment and verification for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2001-01-01

This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

A nuclear criticality safety assessment of the loss of moderation control in 2 1/2 and 10-ton cylinders containing enriched UF6

International Nuclear Information System (INIS)

Newvahner, R.L.; Pryor, W.A.

1991-01-01

Moderation control for maintaining nuclear criticality safety in 2-1/2-ton, 10-ton, and 14-ton cylinders containing enriched uranium hexafluoride (UF 6 ) has been used safely within the nuclear industry for over thirty years, and is dependent on cylinder integrity and containment. This assessment evaluates the loss of moderation control by the breaching of containment and entry of water into the cylinders. The first objective of this study was to estimate the required amounts of water entering these large UF 6 cylinders to react with, and to moderate the uranium compounds sufficiently to cause criticality. Hypothetical accident situations were modeled as a uranyl fluoride (UO 2 F 2 ) slab above a UF 6 hemicylinder, and a UO 2 sphere centered within a UF 6 hemicylinder. These situations were investigated by computational analyses utilizing the KENO V.a Monte Carlo Computer Code. The results were used to estimate both the masses of water required for criticality, and the limiting masses of water that could be considered safe. The second objective of the assessment was to calculate the time available for emergency control actions before a criticality would occur, i.e., a ''safetime,'' for various sources of water and different size openings in a breached cylinder. In the situations considered, except the case for a fire hose, the safetime appears adequate for emergency control actions. The assessment shows that current practices for handling moderation controlled cylinders of low enriched UF 6 , along with the continuation of established personnel training programs, ensure nuclear criticality safety for routine and emergency operations. 2 refs., 5 figs., 1 tab

Safety evaluation for packaging (onsite) product removal can containers

International Nuclear Information System (INIS)

Burnside, M.E.

1998-01-01

Six Product Removal (PR) Cans and Containers are located within the Plutonium Finishing Plant. Each can is expected to contain a maximum of 3 g of residual radioactive material, consisting mainly of plutonium isotopes. The PR Can Containers were previously authorized by HNF-SD-TP-SEP-064, Rev. 0 (Boettger 1997), for the interarea transport of up to 3 g of plutonium. The purpose of this safety evaluation for packaging is to allow the transport of six PR Cans with their Containers from the Plutonium Finishing Plant to the 233 S Evaporator Facility. This safety evaluation for packaging is authorized for use until April 29, 1999, or until the shipment is made, whichever happens first

Safety assessment guidance in the International Atomic Energy Agency RADWASS Program

Energy Technology Data Exchange (ETDEWEB)

Vovk, I.F.; Seitz, R.R.

1995-12-31

The IAEA RADWASS programme is aimed at establishing a coherent and comprehensive set of principles and standards for the safe management of waste and formulating the guidelines necessary for their application. A large portion of this programme has been devoted to safety assessments for various waste management activities. Five Safety Guides are planned to be developed to provide general guidance to enable operators and regulators to develop necessary framework for safety assessment process in accordance with international recommendations. They cover predisposal, near surface disposal, geological disposal, uranium/thorium mining and milling waste, and decommissioning and environmental restoration. The Guide on safety assessment for near surface disposal is at the most advanced stage of preparation. This draft Safety Guide contains guidance on description of the disposal system, development of a conceptual model, identification and description of relevant scenarios and pathways, consequence analysis, presentation of results and confidence building. The set of RADWASS publications is currently undergoing in-depth review to ensure a harmonized approach throughout the Safety Series.

The Nirex safety assessment research programme for 1987/88

International Nuclear Information System (INIS)

Cooper, M.J.; Tasker, P.W.

1987-10-01

This report outlines the work of the Nirex Safety Assessment Research Programme during the period 1st April 1987 to 31st March 1988. The research programme has the specific objective of providing the information requirements of the post-emplacement radiological safety case for the disposal of low-level and intermediate-level radioactive waste in underground repositories. For convenience the programme has been divided into seven areas: physical containment, near-field radionuclide chemistry, evolution of the near-field aqueous environment, mass transfer in the geosphere, the biosphere, gas evolution and migration, and integrated studies. The near-field includes the waste, its immobilising medium, its container, the engineered structure in which the container is emplaced and the immediately adjacent geological formation disturbed by the construction of the repository. (author)

Assessment of Containment Structures Against Missile Impact Threats

Institute of Scientific and Technical Information of China (English)

LI Q M

2006-01-01

In order to ensure the highest safety requirements,nuclear power plant structures (the containment structures,the fuel storages and transportation systems) should be assessed against all possible internal and external impact threats.The internal impact threats include kinetic missiles generated by the failure of high pressure vessels and pipes,the failure of high speed rotating machineries and accidental drops.The external impact threats may come from airborne missiles,aircraft impact,explosion blast and fragments.The impact effects of these threats on concrete and steel structures in a nuclear power plant are discussed.Methods and procedures for the impact assessment of nuclear power plants are introduced.Recent studies on penetration and perforation mechanics as well as progresses on dynamic properties of concrete-like materials are presented to increase the understanding of the impact effects on concrete containment structures.

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Science.gov (United States)

Haslberger, Alexander G

2006-05-03

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

Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

International Nuclear Information System (INIS)

Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

2009-01-01

The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

Prospects for probabilistic safety assessment

International Nuclear Information System (INIS)

Hirschberg, S.

1992-01-01

This article provides some reflections on future developments of Probabilistic Safety Assessment (PSA) in view of the present state of the art and evaluates current trends in the use of PSA for safety management. The main emphasis is on Level 1 PSA, although Level 2 aspects are also highlighted to some extent. As a starting point, the role of PSA is outlined from a historical perspective, demonstrating the rapid expansion of the uses of PSA. In this context the wide spectrum of PSA applications and the associated benefits to the users are in focus. It should be kept in mind, however, that PSA, in spite of its merits, is not a self-standing safety tool. It complements deterministic analysis and thus improves understanding and facilitating prioritization of safety issues. Significant progress in handling PSA limitations - such as reliability data, common-cause failures, human interactions, external events, accident progression, containment performance, and source-term issues - is described. This forms a background for expected future developments of PSA. Among the most important issues on the agenda for the future are PSA scope extensions, methodological improvements and computer code advancements, and full exploitation of the potential benefits of applications to operational safety management. Many PSA uses, if properly exercised, lead to safety improvements as well as major burden reductions. The article provides, in addition, International Atomic Energy Agency (IAEA) perspective on the topics covered, as reflected in the current PSA programs of the agency. 74 refs., 6 figs., 1 tab

The role of risk assessment and safety analysis in integrated safety assessments

International Nuclear Information System (INIS)

Niall, R.; Hunt, M.; Wierman, T.E.

1990-01-01

To ensure that the design and operation of both nuclear and non- nuclear hazardous facilities is acceptable, and meets all societal safety expectations, a rigorous deterministic and probabilistic assessment is necessary. An approach is introduced, founded on the concept of an ''Integrated Safety Assessment.'' It merges the commonly performed safety and risk analyses and uses them in concert to provide decision makers with the necessary depth of understanding to achieve ''adequacy.'' 3 refs., 1 fig

A nuclear criticality safety assessment of the loss of moderation control in 2 1/2 and 10-ton cylinders containing enriched UF{sub 6}

Energy Technology Data Exchange (ETDEWEB)

Newvahner, R.L. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

1991-12-31

Moderation control for maintaining nuclear criticality safety in 2 {1/2}-ton, 10-ton, and 14-ton cylinders containing enriched uranium hexafluoride (UF{sub 6}) has been used safely within the nuclear industry for over thirty years, and is dependent on cylinder integrity and containment. This assessment evaluates the loss of moderation control by the breaching of containment and entry of water into the cylinders. The first objective of this study was to estimate the required amounts of water entering these large UF{sub 6} cylinders to react with, and to moderate the uranium compounds sufficiently to cause criticality. Hypothetical accident situations were modeled as a uranyl fluoride (UO{sub 2}F{sub 2}) slab above a UF{sub 6} hemicylinder, and a UO{sub 2}F{sub 2} sphere centered within a UF{sub 6} hemicylinder. These situations were investigated by computational analyses utilizing the KENO V.a Monte Carlo Computer Code. The results were used to estimate both the masses of water required for criticality, and the limiting masses of water that could be considered safe. The second objective of the assessment was to calculate the time available for emergency control actions before a criticality would occur, i.e., a {open_quotes}safetime{close_quotes}, for various sources of water and different size openings in a breached cylinder. In the situations considered, except the case for a fire hose, the safetime appears adequate for emergency control actions. The assessment shows that current practices for handling moderation controlled cylinders of low enriched UF{sub 6}, along with the continuation of established personnel training programs, ensure nuclear criticality safety for routine and emergency operations.

Human systemic exposure to [¹⁴C]-paraphenylenediamine-containing oxidative hair dyes: Absorption, kinetics, metabolism, excretion and safety assessment.

Science.gov (United States)

Nohynek, Gerhard J; Skare, Julie A; Meuling, Wim J A; Wehmeyer, Kenneth R; de Bie, Albertus Th H J; Vaes, Wouter H J; Dufour, Eric K; Fautz, Rolf; Steiling, Winfried; Bramante, Mario; Toutain, Herve

2015-07-01

Systemic exposure was measured in humans after hair dyeing with oxidative hair dyes containing 2.0% (A) or 1.0% (B) [(14)C]-p-phenylenediamine (PPD). Hair was dyed, rinsed, dried, clipped and shaved; blood and urine samples were collected for 48 hours after application. [(14)C] was measured in all materials, rinsing water, hair, plasma, urine and skin strips. Plasma and urine were also analysed by HLPC/MS/MS for PPD and its metabolites (B). Total mean recovery of radioactivity was 94.30% (A) or 96.21% (B). Mean plasma Cmax values were 132.6 or 97.4 ng [(14)C]-PPDeq/mL, mean AUC(0-∞) values 1415 or 966 ng [(14)C]-PPDeq/mL*hr in studies A or B, respectively. Urinary excretion of [(14)C] mainly occurred within 24 hrs after hair colouring with a total excretion of 0.72 or 0.88% of applied radioactivity in studies A or B, respectively. Only N,N'-diacetylated-PPD was detected in plasma and the urine. A TK-based human safety assessment estimated margins of safety of 23.3- or 65-fold relative to respective plasma AUC or Cmax values in rats at the NOAEL of a toxicity study. Overall, hair dyes containing PPD are unlikely to pose a health risk since they are used intermittently and systemic exposure is limited to the detoxified metabolite N,N'-diacetyl-PPD. Copyright © 2015 Elsevier Ltd. All rights reserved.

Determination of the number of software tests using probabilistic safety assessment

International Nuclear Information System (INIS)

Kang, H. K.; Seong, T. Y.; Lee, K. Y.

2000-01-01

The broader usage of digital equipment in nuclear power plants gives rise to the safety problems of software. The field test should be performed before the software is used in critical applications because it is well known that software shows non-linear response when it is applied to different target systems in different environment. In the case of safety-critical applications, the result of tests contains usually zero failure case and the satisfiable number of tests is hard to be determined. In this paper, we suggests the method to determine the number of software tests without failure using the probabilistic safety assessment. From the result of the probabilistic safety assessment on total system, the desirable unavailability of software is calculated and the number of tests is determined

Safety and efficacy of personal care products containing colloidal oatmeal

Directory of Open Access Journals (Sweden)

Criquet M

2012-11-01

Full Text Available Maryline Criquet,1 Romain Roure,1 Liliane Dayan,2 Virginie Nollent,1 Christiane Bertin11Johnson & Johnson Santé Beauté France, Issy les Moulineaux, 2Independent consultant dermatologist, Paris, FranceBackground: Colloidal oatmeal is a natural ingredient used in the formulation of a range of personal care products for relief of skin dryness and itchiness. It is also used as an adjunctive product in atopic dermatitis. The safety of personal care products used on vulnerable skin is of particular importance and the risk of developing further skin irritations and/or allergies should be minimized.Methods: In a series of studies, we tested the safety of personal care products containing oatmeal (creams, cleansers, lotions by assessing their irritant/allergenic potential on repeat insult patch testing, in safety-in-use and ocular studies using subjects with nonsensitive and sensitive skin. We also tested the skin moisturizing and repair properties of an oatmeal-containing skin care product for dry skin.Results: We found that oatmeal-containing personal care products had very low irritant potential as well as a very low allergenic sensitization potential. Low-level reactions were documented in 1.0% of subjects during the induction phase of repeat insult patch testing; one of 2291 subjects developed a persistent but doubtful low-level reaction involving edema during the challenge phase in repeat insult patch testing. No allergies were reported by 80 subjects after patch testing after in-use application. Sustained skin moisturizing was documented in subjects with dry skin that lasted up to 2 weeks after product discontinuation.Conclusion: Our results demonstrate that colloidal oatmeal is a safe and effective ingredient in personal care products. No allergies were reported by consumers of 445,820 products sold during a 3-year period.Keywords: Avena sativa, colloids, protective agents, atopic dermatitis, irritant dermatitis, allergenic dermatitis, skin

Improvement of operational safety: The self-assessment at the Russian NPPs

International Nuclear Information System (INIS)

Kolotov, Aleksander

2002-01-01

The operating organization has scheduled for the forthcoming year to elaborate the NPP safety self-assessment standards and to settle precise criterion for its performance. Toward this end it was decided to form a Working Team including of NPP and VNIIAES representatives to elaborate major tasks on the self-assessment improvement, development and implementation of new documentation and training the personnel in new methodology. Actions developed by R osenergoatom , one of its items is the participation of 'Rosenergoatom' and VNIIAES representatives in the workshop, contains the sequence of Russian NPP safety self-assessment improvement at the first stage

Safety assessment of HLW geological disposal system

International Nuclear Information System (INIS)

Naito, Morimasa

2006-01-01

In accordance with the Japanese nuclear program, the liquid waste with a high level of radioactivity arising from reprocessing is solidified in a stable glass matrix (vitrification) in stainless steel fabrication containers. The vitrified waste is referred to as high-level radioactive waste (HLW), and is characterized by very high initial radioactivity which, even though it decreases with time, presents a potential long-term risk. It is therefore necessary to thoroughly manage HLW from human and his environment. After vitrification, HLW is stored for a period of 30 to 50 years to allow cooling, and finally disposed of in a stable geological environment at depths greater than 300 m below surface. The deep underground environment, in general, is considered to be stable over geological timescales compared with surface environment. By selecting an appropriate disposal site, therefore, it is considered to be feasible to isolate the waste in the repository from man and his environment until such time as radioactivity levels have decayed to insignificance. The concept of geological disposal in Japan is similar to that in other countries, being based on a multibarrier system which combines the natural geological environment with engineered barriers. It should be noted that geological disposal concept is based on a passive safety system that does not require any institutional control for assuring long term environmental safety. To demonstrate feasibility of safe HLW repository concept in Japan, following technical steps are essential. Selection of a geological environment which is sufficiently stable for disposal (site selection). Design and installation of the engineered barrier system in a stable geological environment (engineering measures). Confirmation of the safety of the constructed geological disposal system (safety assessment). For site selection, particular consideration is given to the long-term stability of the geological environment taking into account the fact

Assessment of Safety Culture

International Nuclear Information System (INIS)

Bilic Zabric, T.; Kavsek, D.

2006-01-01

A strong safety culture leads to more effective conduct of work and a sense of accountability among managers and employees, who should be given the opportunity to expand skills by training. The resources expended would thus result in tangible improvements in working practices and skills, which encourage further improvement of safety culture. In promoting an improved safety culture, NEK has emphasized both national and organizational culture with an appropriate balance of behavioural sciences and quality management systems approaches. In recent years there has been particular emphasis put on an increasing awareness of the contribution that human behavioural sciences can make to develop good safety practices. The purpose of an assessment of safety culture is to increase the awareness of the present culture, to serve as a basis for improvement and to keep track of the effects of change or improvement over a longer period of time. There is, however, no single approach that is suitable for all purposes and which can measure, simultaneously, all the intangible aspects of safety culture, i.e. the norms, values, beliefs, attitudes or the behaviours reflecting the culture. Various methods have their strengths and weaknesses. To prevent significant performance problems, self-assessment is used. Self-assessment is the process of identifying opportunities for improvement actively or, in some cases, weaknesses that could cause more serious errors or events. Self-assessments are an important input to the corrective action programme. NEK has developed questionnaires for safety culture self-assessment to obtain information that is representative of the whole organization. Questionnaires ensure a greater degree of anonymity, and create a less stressful situation for the respondent. Answers to questions represent the more apparent and conscious values and attitudes of the respondent. NEK proactively co-operates with WANO, INPO, IAEA in the areas of Safety Culture and Human

A Framework for Assessment of Aviation Safety Technology Portfolios

Science.gov (United States)

Jones, Sharon M.; Reveley, Mary S.

2014-01-01

The programs within NASA's Aeronautics Research Mission Directorate (ARMD) conduct research and development to improve the national air transportation system so that Americans can travel as safely as possible. NASA aviation safety systems analysis personnel support various levels of ARMD management in their fulfillment of system analysis and technology prioritization as defined in the agency's program and project requirements. This paper provides a framework for the assessment of aviation safety research and technology portfolios that includes metrics such as projected impact on current and future safety, technical development risk and implementation risk. The paper also contains methods for presenting portfolio analysis and aviation safety Bayesian Belief Network (BBN) output results to management using bubble charts and quantitative decision analysis techniques.

Mark I containment, short term program. Safety evaluation report

International Nuclear Information System (INIS)

1977-12-01

Presented is a Safety Evaluation Report (SER) prepared by the Office of Nuclear Reactor Regulation addressing the Short Term Program (STP) reassessment of the containment systems of operating Boiler Water Reactor (BWR) facilities with the Mark I containment system design. The information presented in this SER establishes the basis for the NRC staff's conclusion that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety of the public, during an interim period of approximately two years while a methodical, comprehensive Long Term Program (LTP) is conducted. This SER also provides one of the basic foundations for the NRC staff review of the Mark I containment systems for facilities not yet licensed for operation

Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Baji, C.; Vamos, G.; Toth, J.

2001-01-01

The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

Analysis of safety margins for PuO2 containers

International Nuclear Information System (INIS)

Hubert, P.; Tomachevsky, E.

1987-11-01

In the regular manner the containers for PuO 2 transport are type B(U) and give satisfaction to the AIEA proofs. However the vigour of this conception's containers and the analysis of other radioactive containers permit to think that large safety margins exist. In this paper, the importance and the kind of these margins are studied [fr

Nirex Safety Assessment Research Programme bibliography, 1988

International Nuclear Information System (INIS)

Cooper, M.J.

1988-05-01

This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. All work referred to has been funded, or partly funded, by UK Nirex Limited, previously known as the Nuclear Industry Radioactive Waste Executive (NIREX). The bibliography has been divided into two sections, a List of Publications in roughly chronological order and an Author Index. The topics involved include near-field and far-field studies. The near-field includes the waste, its immobilising medium, its container, the engineered structure in which the containers are emplaced, and the immediately adjacent geological formation disturbed by the construction of the repository. The far-field is the undisturbed geological formation between the near-field and the biosphere. (author)

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

Based on several Safety Culture self-assessment applications in nuclear organisations, the paper stresses relevant aspects to be considered when programming an assessment of this type. Reasons for assessing Safety Culture, basic principles to take into account, necessary resources, the importance of proper statistical analyses, the feed-back of results, and the setting up of action plans to enhance Safety Culture are discussed. (author)

Flamanville 3 EPR, Safety Assessment and On-site Inspections

International Nuclear Information System (INIS)

Piedagnel, Corinne; Tarallo, Francois; Monnot, Bernard

2011-01-01

As a Technical Support Organisation of the French Safety Authority (ASN), the IRSN carries out the safety assessment of EPR project design and participates in the ASN inspections performed at the construction site and in factories. The design assessment consists in defining the safety functions which should be ensured by civil structures, evaluating the EPR Technical Code for Civil works (ETC-C) in which EdF has defined design criteria and construction rules, and carrying out a detailed assessment of a selection of safety-related structures. Those detailed assessments do not consist of a technical control but of an analysis whose objectives are to ensure that design and demonstrations are robust, in accordance with safety and regulatory rules. Most assessments led IRSN to ask EdF to provide additional justification sometimes involving significant modifications. In the light of those complementary justifications and modifications, IRSN concluded that assessments carried out on design studies were globally satisfactory. The participation of IRSN to the on-site inspections led by ASN is a part of the global control of the compliance of the reactor with its safety objectives. For that purpose IRSN has defined a methodology and an inspection program intended to ASN: based on safety functions associated with civil works (confinement and resistance to aggressions), the corresponding behaviour requirements are identified and linked to a list of main civil works elements. During the inspections, deviations to the project's technical specifications or to the rules of the art were pointed out by IRSN. Those deviations cover various items, such as concrete fabrication, concrete pouring methodology, lack of reinforcement in some structures, unadapted welding procedures of the containment leak-tight steel liner and unsatisfactory treatment of concreting joints. The analysis of those problems has revealed flaws in the organisation of the contractors teams together with an

Developing guidance in the nuclear criticality safety assessment for fuel cycle facilities

International Nuclear Information System (INIS)

Galet, C.; Evo, S.

2012-01-01

In this poster IRSN (Institute for radiation protection and nuclear safety) presents its safety guides whose purpose is to transmit the safety assessment know-how to any 'junior' staff or even to give a view of the safety approach on the overall risks to any staff member. IRSN has written a first version of such a safety guide for fuel cycle facilities and laboratories. It is organized into several chapters: some refer to types of assessments, others concern the types of risks. Currently, this guide contains 13 chapters and each chapter consists of three parts. In parallel to the development of criticality chapter of this guide, the IRSN criticality department has developed a nuclear criticality safety guide. It follows the structure of the three parts fore-mentioned, but it presents a more detailed first part and integrates, in the third part, the experience feedback collected on nuclear facilities. The nuclear criticality safety guide is online on the IRSN's web site

OSART Independent Safety Culture Assessment (ISCA) Guidelines

International Nuclear Information System (INIS)

2016-01-01

Safety culture is understood as an important part of nuclear safety performance. This has been demonstrated by the analysis of significant events such as Chernobyl, Davis Besse, Vandellos II, Asco, Paks, Mihamma and Forsmark, among others. In order to enhance safety culture, one essential activity is to perform assessments. IAEA Safety Standard Series No. GS-R-3, The Management System for Facilitites and Activities, states requirements for continuous improvement of safety culture, of which self, peer and independent safety culture assessments constitute an essential part. In line with this requirement, the Independent Safety Culture Assessment (ISCA) module is offered as an add-on module to the IAEA Operational Safety Review Team (OSART) programme. The OSART programme provides advice and assistance to Member States to enhance the safety of nuclear power plants during commissioning and operation. By including the ISCA module in an OSART mission, the receiving organization benefits from the synergy between the technical and the safety culture aspects of the safety review. The joint operational safety and safety culture assessment provides the organization with the opportunity to better understand the interactions between technical, human, organizational and cultural aspects, helping the organization to take a systemic approach to safety through identifying actions that fully address the root causes of any identified issue. Safety culture assessments provide insight into the fundamental drivers that shape organizational patterns of behaviour, safety consciousness and safety performance. The complex nature of safety culture means that the analysis of the results of such assessments is not as straightforward as for other types of assessment. The benefits of the results of nuclear safety culture assessments are maximized only if appropriate tools and guidance for these assessments is used; hence, this comprehensive guideline has been developed. The methodology explained

Two types of a passive safety containment for a near future BWR with active and passive safety systems

International Nuclear Information System (INIS)

Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

2009-01-01

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

Two types of a passive safety containment for a near future BWR with active and passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Sato, Takashi [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)], E-mail: takashi44.sato@glb.toshiba.co.jp; Akinaga, Makoto; Kojima, Yoshihiro [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)

2009-09-15

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

Seismic safety margin assessment program (Annual safety research report, JFY 2010)

International Nuclear Information System (INIS)

Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

2011-01-01

Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

Metrics design for safety assessment

NARCIS (Netherlands)

Luo, Yaping; van den Brand, M.G.J.

2016-01-01

Context:In the safety domain, safety assessment is used to show that safety-critical systems meet the required safety objectives. This process is also referred to as safety assurance and certification. During this procedure, safety standards are used as development guidelines to keep the risk at an

The Fukushima Daiichi Accident. Technical Volume 2/5. Safety Assessment

International Nuclear Information System (INIS)

2015-08-01

Technical Volume 1 of this report has described what happened during the accident at the Fukushima Daiichi nuclear power plant (NPP). This volume begins (Section 2.1) with a review of how the design basis of the site for external events was assessed initially and then reassessed over the life of the NPP. The section also describes the physical changes that were made to the units as a result. The remainder of the volume describes the treatment of beyond design basis events in the safety assessment of the site, the accident management provisions, the effectiveness of regulatory programmes, human and organizational factors and the safety culture, and the role of operating experience. Further background information is contained in three annexes included on the CD-ROM of this Technical Volume which describe analytical investigations of the accident along with information on topics such as system performance, defence in depth and severe accident phenomena. Section 2.2 provides an assessment of the systems that failed, resulting in a failure to maintain the fundamental safety functions in Units 1–3, which were in operation at the time of the tsunami and in which the reactor pressure vessels (RPV) and containment vessels failed. The section also describes Units 4-6, which were shut down at the time of the tsunami, and the site’s central spent fuel storage facility. Section 2.3 discusses the probabilistic and deterministic safety assessments of beyond design basis accidents (BDBAs) that had been performed for the plant and the insights from these assessments that had led to changes in the plant’s design. The section pays particular attention to the assessment of extreme natural hazards, such as the one which led to the total loss of AC power supply on the site. The additional loss of DC power supply in Units 1 and 2 played a key role in the progression of the accident because it impeded the diagnosis of plant conditions and made the operators unaware of the status of

Safety assessment principles for nuclear plants

International Nuclear Information System (INIS)

1992-01-01

The present Safety Assessment Principles result from the revision of those which were drawn up following a recommendation arising from the Sizewell-B enquiry. The principles presented here relate only to nuclear safety; there is a section on risks from normal operation and accident conditions and the standards against which those risks are assessed. A major part of the document deals with the principles that cover the design of nuclear plants. The revised Safety assessment principles are aimed primarily at the safety assessment of new nuclear plants but they will also be used in assessing existing plants. (UK)

Development of safety related technology and infrastructure for safety assessment

International Nuclear Information System (INIS)

Venkat Raj, V.

1997-01-01

Development and optimum utilisation of any technology calls for the building up of the necessary infrastructure and backup facilities. This is particularly true for a developing country like India and more so for an advanced technology like nuclear technology. Right from the inception of its nuclear power programme, the Indian approach has been to develop adequate infrastructure in various areas such as design, construction, manufacture, installation, commissioning and safety assessment of nuclear plants. This paper deals with the development of safety related technology and the relevant infrastructure for safety assessment. A number of computer codes for safety assessment have been developed or adapted in the areas of thermal hydraulics, structural dynamics etc. These codes have undergone extensive validation through data generated in the experimental facilities set up in India as well as participation in international standard problem exercises. Side by side with the development of the tools for safety assessment, the development of safety related technology was also given equal importance. Many of the technologies required for the inspection, ageing assessment and estimation of the residual life of various components and equipment, particularly those having a bearing on safety, were developed. This paper highlights, briefly, the work carried out in some of the areas mentioned above. (author)

Thinking of the safety assessment of HLW disposal

International Nuclear Information System (INIS)

Li Honghui; Zhao Shuaiwei; Liu Jianqin; Liu Wei; Wan Lei; Yang Zhongtian; An Hongxiang; Sun Qinghong

2014-01-01

The function and the research methods of safety assessment are discussed. Two methods about safety assessment and the requirement of safety assessment are introduced. The key parameters and influence factors in nuclide transport of safety assessment are specialized. The works will be done on safety assessment is discussed which will give some suggests for the development of safety assessment. (authors)

ALARP considerations in criticality safety assessments

International Nuclear Information System (INIS)

Bowden, Russell L.; Barnes, Andrew; Thorne, Peter R.; Venner, Jack

2003-01-01

Demonstrating that the risk to the public and workers is As Low As Reasonably Practicable (ALARP) is a fundamental requirement of safety cases for nuclear facilities in the United Kingdom. This is embodied in the Safety Assessment Principles (SAPs) published by the Regulator, the essence of which is incorporated within the safety assessment processes of the various nuclear site licensees. The concept of ALARP within criticality safety assessments has taken some time to establish in the United Kingdom. In principle, the licensee is obliged to search for a deterministic criticality safety solution, such as safe geometry vessels and passive control features, rather than placing reliance on active measurement devices and plant administrative controls. This paper presents a consideration of some ALARP issues in relation to the development of criticality safety cases. The paper utilises some idealised examples covering a range of issues facing the criticality safety assessor, including new plant design, operational plant and decommissioning activities. These examples are used to outline the elements of the criticality safety cases and present a discussion of ALARP in the context of criticality safety assessments. (author)

Safety culture assessment developed by JANTI

International Nuclear Information System (INIS)

Hamada, Jun

2009-01-01

Japan's JCO accident in September 1999 provided a real-life example of what can happen when insufficient attention is paid to safety culture. This accident brought to light the importance of safety culture and reinforced the movement to foster a safety culture. Despite this, accidents and inappropriate conduct have continued to occur. Therefore, there is a strong demand to instill a safety culture throughout the nuclear power industry. In this context, Japan's nuclear power regulator, the Nuclear and Industrial Safety Agency (NISA), decided to include in its safety inspections assessments of the safety culture found in power utilities' routine safety operations to get signs of deterioration in the organizational climate. In 2007, NISA constructed guidelines for their inspectors to carry out these assessments. At the same time, utilities have embarked on their own independent safety culture initiatives, such as revising their technical specifications and building effective PDCA cycle to promote safety culture. In concert with these developments, JANTI has also instituted safety culture assessments. (author)

Rad waste disposal safety analysis / Integrated safety assessment of a waste repository

International Nuclear Information System (INIS)

Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

2012-04-01

We developed CYPRUS+and adopted PID and RES method for the development of scenario. Safety performance assessment program was developed using GoldSim for the safety assessment of disposal system for the disposal of spnet fuels and wastes resulting from the pyrpoprocessing. Biosphere model was developed and verified in cooperation with JAEA. The capability to evaluate post-closure performance and safety was added to the previously developed program. And, nuclide migration and release to the biosphere considering site characteristics was evaluated by using deterministic and probabilistic approach. Operational safety assessment for drop, fire, and earthquake was also statistically evaluated considering well-established input parameter distribution. Conservative assessment showed that dose rate is below the limit value of low- and intermediate-level repository. Gas generation mechanism within engineered barrier was defined and its influence on safety was evaluated. We made probabilistic safety assessment by obtaining the probability distribution functions of important input variables and also made a sensitivity analysis. The maximum annual dose rate was shown to be below the safety limit value of 10 mSv/yr. The structure and element of safety case was developed to increase reliability of safety assessment methodology for a deep geological repository. Finally, milestone for safety case development and implementation strategy for each safety case element was also proposed

Probabilistic safety assessment for Hanford high-level waste tanks

International Nuclear Information System (INIS)

MacFarlane, D.R.; Stack, D.S.; Kindinger, J.P.; Deremer, R.K.

1995-01-01

This paper gives results from the first comprehensive level-3 probabilistic safety assessment (PSA), including consideration of external events, for the Hanford tank farm (HTF). This work was sponsored by the U.S. Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM). At the HTF, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/saltcake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is ∼60 million gal, containing ∼200 million Ci of radioactivity

Safety and risk assessment of ceramide 3 in cosmetic products.

Science.gov (United States)

Choi, Seul Min; Lee, Byung-Mu

2015-10-01

Ceramide 3 is used mainly as a moisturizer in various cosmetic products. Although several safety studies on formulations containing pseudo-ceramide or ceramide have been conducted at the preclinical and clinical levels for regulatory approval, no studies have evaluated the systemic toxicity of ceramide 3. To address this issue, we conducted a risk assessment and comprehensive toxicological review of ceramide and pseudo-ceramide. We assumed that ceramide 3 is present in various personal and cosmetic products at concentrations of 0.5-10%. Based on previously reported exposure data, the margin of safety (MOS) was calculated for product type, use pattern, and ceramide 3 concentration. Lipsticks with up to 10% ceramide 3 (MOS = 4111) are considered safe, while shampoos containing 0.5% ceramide 3 (MOS = 148) are known to be safe. Reported MOS values for body lotion applied to the hands (1% ceramide 3) and back (5% ceramide 3) were 103 and 168, respectively. We anticipate that face cream would be safe up to a ceramide 3 concentration of 3% (MOS = 149). Collectively, the MOS approach indicated no safety concerns for cosmetic products containing less than 1% ceramide 3. Copyright © 2015 Elsevier Ltd. All rights reserved.

Immunization Safety Review: Thimerosal - Containing Vaccines and Neurodevelopmental Disorders

National Research Council Canada - National Science Library

Stratton, Kathleen; Gable, Alicia; McCormick, Marie C

2001-01-01

In this report, the Immunization Safety Review committee examines the hypothesis of whether or not the use of vaccines containing the preservative thimerosal can cause neurodevelopmental disorders (NDDs...

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

International Nuclear Information System (INIS)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-01

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments that lie outside the

Probabilistic safety assessment of nuclear power plants: a monograph

International Nuclear Information System (INIS)

Solanki, R.B.; Prasad, Mahendra

2007-11-01

This monograph on probabilistic safety assessment (PSA) is addressed to the wide community of professionals engaged in the nuclear industry and concerned with the safety issues of nuclear power plants (NPPs). While the monograph describes PSA of NPPs, the principles described in this monograph can be extended to other facilities like spent fuel storage, fuel reprocessing plants and non-nuclear facilities like chemical plants, refineries etc. as applicable. The methodology for risk assessment in chemical plants or refineries is generally known as quantitative risk analysis (QRA). The fundamental difference between NPP and chemical plant is that in NPPs the hazardous material (fuel and fission products) are contained at a single location (i.e. inside containment), whereas in a chemical plant and reprocessing plants, the hazardous material is present simultaneously at many places, like pipelines, reaction towers, storage tanks, etc. Also unlike PSA, QRA does not deal with levels; it uses an integrated approach combining all the levels. The monograph covers the areas of broad interest in the field of PSA such as historical perspective, fundamentals of PSA, strengths and weaknesses of PSA, applications of PSA, role of PSA in the regulatory decision making and issues for advancement of PSA

The Safety Case and Safety Assessment for the Disposal of Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-09-15

This Safety Guide provides guidance and recommendations on meeting the safety requirements in respect of the safety case and supporting safety assessment for the disposal of radioactive waste. The safety case and supporting safety assessment provide the basis for demonstration of safety and for licensing of radioactive waste disposal facilities and assist and guide decisions on siting, design and operations. The safety case is also the main basis on which dialogue with interested parties is conducted and on which confidence in the safety of the disposal facility is developed. This Safety Guide is relevant for operating organizations preparing the safety case as well as for the regulatory body responsible for developing the regulations and regulatory guidance that determine the basis and scope of the safety case. Contents: 1. Introduction; 2. Demonstrating the safety of radioactive waste disposal; 3. Safety principles and safety requirements; 4. The safety case for disposal of radioactive waste; 5. Radiological impact assessment for the period after closure; 6. Specific issues; 7. Documentation and use of the safety case; 8. Regulatory review process.

Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2

International Nuclear Information System (INIS)

Osborn, R.N.; Olson, J.; Sommers, P.E.

1983-08-01

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators

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 Culture Monitoring: How to Assess Safety Culture in Real Time?

International Nuclear Information System (INIS)

Zronek, B.; Maryska, J.; Treslova, L.

2016-01-01

Do you know what is current level of safety culture in your company? Are you able to follow trend changes? Do you know what your recent issues are? Since safety culture is understood as vital part of nuclear industry daily life, it is crucial to know what the current level is. It is common to perform safety culture survey or ad hoc assessment. This contribution shares Temelin NPP, CEZ approach how to assess safety culture level permanently. Using behavioral related outputs of gap solving system, observation program, dedicated surveys, regulatory assessment, etc., allows creating real time safety culture monitoring without the need to perform any other activities. (author)

Regulatory review of safety cases and safety assessments - associated challenges

International Nuclear Information System (INIS)

Bennett, D.G.; Ben Belfadhel, M.; Metcalf, P.E.

2006-01-01

Regulatory reviews of safety cases and safety assessments are essential for credible decision making on the licensing or authorization of radioactive waste disposal facilities. Regulatory review also plays an important role in developing the safety case and in establishing stakeholders' confidence in the safety of the facility. Reviews of safety cases for radioactive waste disposal facilities need to be conducted by suitably qualified and experienced staff, following systematic and well planned review processes. Regulatory reviews should be sufficiently comprehensive in their coverage of issues potentially affecting the safety of the disposal system, and should assess the safety case against clearly established criteria. The conclusions drawn from a regulatory review, and the rationale for them should be reproducible and documented in a transparent and traceable way. Many challenges are faced when conducting regulatory reviews of safety cases. Some of these relate to issues of project and programme management, and resources, while others derive from the inherent difficulties of assessing the potential long term future behaviour of engineered and environmental systems. The paper describes approaches to the conduct of regulatory reviews and discusses some of the challenges faced. (author)

Serviceability design load factors and reliability assessments for reinforced concrete containment structures

International Nuclear Information System (INIS)

Han Bong Koo

1998-01-01

A reinforced concrete nuclear power plant containment structure is subjected to various random static and stochastic loads during its lifetime. Since these loads involve inherent randomness and other uncertainties, an appropriate probabilistic model for each load must be established in order to perform reliability analysis. The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops probability-based load factors for the limit state design of reinforced concrete containment structures. The purpose of constructing reinforced concrete containment structure is to protect against radioactive release, and so the use of a serviceability limit state against crack failure that can cause the emission of radioactive materials is suggested as a critical limit state for reinforced concrete containment structures. Load factors for the design of reinforced concrete containment structures are proposed and carried out the reliability assessments. (orig.)

Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

Safety assessments for potential exposures

International Nuclear Information System (INIS)

Dunn, D.I.

2012-04-01

Safety Assessment of potential exposures have been carried out in major practices, namely: industrial radiography, gamma irradiators and electron accelerators used in industry and research, and radiotherapy. This paper focuses on reviewing safety assessment methodologies and using developed software to analyse radiological accidents, also review, and discuss these past accidents.The primary objective of the assessment is to assess the adequacy of planned or existing measures for protection and safety and to identify any additional measures that should be put in place. As such, both routine use of the source and the probability and magnitude of potential exposures arising from accidents or incidents should be considered. Where the assessment indicates that there is a realistic possibility of an accident affecting workers or members of the public or having consequences for the environment, the registrant or licensee should prepare a suitable emergency plan. A safety assessment for normal operation addresses all the conditions under which the radiation source operates as expected, including all phases of the lifetime of the source. Due account needs to be taken of the different factors and conditions that will apply during non-operational phases, such as installation, commissioning and maintenance. (author)

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

International Nuclear Information System (INIS)

Song, Tae Young

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

Safety assessment for spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

Nuclear data for radiation damage assessment and related safety aspects

International Nuclear Information System (INIS)

Kocherov, N.P.

1989-12-01

The IAEA Advisory Group Meeting on Nuclear Data for Radiation Damage Assessment and Related Safety Aspects was held at the IAEA Headquarters in Vienna, 19-22 September 1989. This report contains the conclusions and recommendations of this meeting. The papers which the participants prepared for and presented at the meeting will be published as an IAEA Technical Document. (author)

76 FR 75450 - Safety Zone; Container Crane Relocation, Cooper and Wando Rivers, Charleston, SC

Science.gov (United States)

2011-12-02

...-AA00 Safety Zone; Container Crane Relocation, Cooper and Wando Rivers, Charleston, SC AGENCY: Coast... moving safety zone around a barge transporting two container cranes on the Cooper and Wando Rivers during... barge will transit the Cooper and Wando Rivers. Because of the size of the two container cranes and the...

Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

Energy Technology Data Exchange (ETDEWEB)

1980-10-01

The Assistant Secretary for Environment has responsibility for identifying, characterizing, and ameliorating the environmental, health, and safety issues and public concerns associated with commercial operation of specific energy systems. The need for developing a safety and environmental control assessment for liquefied gaseous fuels was identified by the Environmental and Safety Engineering Division as a result of discussions with various governmental, industry, and academic persons having expertise with respect to the particular materials involved: liquefied natural gas, liquefied petroleum gas, hydrogen, and anhydrous ammonia. This document is arranged in three volumes and reports on progress in the Liquefied Gaseous Fuels (LGF) Safety and Environmental Control Assessment Program made in Fiscal Year (FY)-1979 and early FY-1980. Volume 1 (Executive Summary) describes the background, purpose and organization of the LGF Program and contains summaries of the 25 reports presented in Volumes 2 and 3. Annotated bibliographies on Liquefied Natural Gas (LNG) Safety and Environmental Control Research and on Fire Safety and Hazards of Liquefied Petroleum Gas (LPG) are included in Volume 1.

Process and device for reducing the presence in the safety containment of a nuclear reactor

International Nuclear Information System (INIS)

Stiefel, M.

1981-01-01

If faults occur, the excess pressure in the safety containment has to be reduced. Part of the gas contents of the safety containment is drawn off for this purpose. Hydrogen up to a maximum of 3.5% by volume is added to the drawn-off gas. The gas enriched with hydrogen is taken through a recombiner, where part of the oxygen content of the gas burns with the hydrogen, producing water. The gas with reduced oxygen content is returned to the safety containment via the free end of an annular pipe. The pressure existing in the safety containment is reduced according to the reduction in oxygen content. (orig./HP) [de

Coupled analysis of passive safety injection and containment filtered venting for passive decay heat removal - 15140

International Nuclear Information System (INIS)

Kim, S.H.; Ham, J.H.; Jeong, Y.H.; Chang, S.H.

2015-01-01

Lots of interests for the safety of nuclear power plants have risen these days. The safety has to be continuously reviewed and enhanced in nuclear power plants currently operating as well as those designed and constructed in future. After the Fukushima accidents, many additional safety systems which can be applied to nuclear power plants in operation have been proposed. Those include alternating power source such as movable diesel generators and DC batteries in non-safety grade. Also, emergency preparedness for the prevention of a core damage accident was proposed to cope with the extended-SBO (station blackout) by using fire protection systems. In order to prevent the release of radioactive materials, safety systems for preserving the integrity of containment were proposed in two views of cooling and venting containment. Two approaches are effective for mitigating a severe accident. The design concept installing big water tanks besides containment at high level was proposed for various safety functions. One of the functions in the system is to inject the coolant from the elevated tank into a reactor vessel in the case of loss of coolant accident. When the pressure in reactor coolant system is sufficiently low, the coolant can be injected by gravity. If not, the depressurization in reactor vessel would be needed considering the containment pressure. Containment cooling in conventional pressurized water reactors is dependent on containment cooling pumps and sprays. Additional containment cooling systems cannot be simply and easily applied in the current nuclear power plants without major modifications. Therefore, for the operation of passive safety injection system, containment filtered venting system can be adopted for the depressurization of containment. In the design and operation of the passive safety injection system and the containment filtered venting system, main operating points related with open and close pressures in the filtered venting system were

Consideration of aging in probabilistic safety assessment

International Nuclear Information System (INIS)

Titina, B.; Cepin, M.

2007-01-01

Probabilistic safety assessment is a standardised tool for assessment of safety of nuclear power plants. It is a complement to the safety analyses. Standard probabilistic models of safety equipment assume component failure rate as a constant. Ageing of systems, structures and components can theoretically be included in new age-dependent probabilistic safety assessment, which generally causes the failure rate to be a function of age. New age-dependent probabilistic safety assessment models, which offer explicit calculation of the ageing effects, are developed. Several groups of components are considered which require their unique models: e.g. operating components e.g. stand-by components. The developed models on the component level are inserted into the models of the probabilistic safety assessment in order that the ageing effects are evaluated for complete systems. The preliminary results show that the lack of necessary data for consideration of ageing causes highly uncertain models and consequently the results. (author)

Qualitative uncertainty analysis in probabilistic safety assessment context

International Nuclear Information System (INIS)

Apostol, M.; Constantin, M; Turcu, I.

2007-01-01

In Probabilistic Safety Assessment (PSA) context, an uncertainty analysis is performed either to estimate the uncertainty in the final results (the risk to public health and safety) or to estimate the uncertainty in some intermediate quantities (the core damage frequency, the radionuclide release frequency or fatality frequency). The identification and evaluation of uncertainty are important tasks because they afford credit to the results and help in the decision-making process. Uncertainty analysis can be performed qualitatively or quantitatively. This paper performs a preliminary qualitative uncertainty analysis, by identification of major uncertainty in PSA level 1- level 2 interface and in the other two major procedural steps of a level 2 PSA i.e. the analysis of accident progression and of the containment and analysis of source term for severe accidents. One should mention that a level 2 PSA for a Nuclear Power Plant (NPP) involves the evaluation and quantification of the mechanisms, amount and probabilities of subsequent radioactive material releases from the containment. According to NUREG 1150, an important task in source term analysis is fission products transport analysis. The uncertainties related to the isotopes distribution in CANDU NPP primary circuit and isotopes' masses transferred in the containment, using SOPHAEROS module from ASTEC computer code will be also presented. (authors)

A Methodology for Safety Culture Impact Assessment

Energy Technology Data Exchange (ETDEWEB)

Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

2014-05-15

The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively.

A Methodology for Safety Culture Impact Assessment

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2014-01-01

The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

Energy Technology Data Exchange (ETDEWEB)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-15

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments

SKI's and SSI's joint review of SKB's safety assessment report, SR 97. Summary

International Nuclear Information System (INIS)

2001-01-01

The Swedish Nuclear Fuel and Waste Management Co (SKB) has a programme for the siting of a repository for spent nuclear fuel in Swedish bedrock. In 1996, the Swedish Government decided that SKB must perform an assessment of the repository's long-term safety before undertaking the next step of the programme which entails drilling in a minimum of two municipalities (site investigations). SKB has presented such a safety assessment in SR 97 Post-closure Safety (henceforth referred to as SR 97). SR 97 is one of the documents in the comprehensive reporting that SKB must provide when it proposes sites for investigation. The Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Institute (SSI) have evaluated SR 97 in terms of its purposes which are to demonstrate a methodology for safety assessment, to show that Swedish bedrock can provide a safe repository using SKB's method, to provide a basis for specifying the factors that are important for site selection and to derive preliminary requirements on the function of the engineered barriers. The authorities have reached the following conclusions: SR 97 does not indicate any conditions that would mean that geological final disposal in accordance with SKB's method would have significant deficiencies in relation to the safety and radiation protection requirements of the authorities. SR 97 contains the elements required for a comprehensive assessment of safety and radiation protection. SKB's safety assessment methodology has improved within several important areas, such as the documentation of processes and properties that can affect repository performance and the development of models for safety assessment calculations. The methodology used in SR 97 has some deficiencies, for example, the specification of future events to be described in the safety assessment. SR 97 has not, to an adequate extent, dealt with unfavourable conditions that can affect the future safety of a repository. SKB states that the

77 FR 9847 - Safety Zone; Kinnickinnic River Containment and Cleanup; Milwaukee, WI

Science.gov (United States)

2012-02-21

... Kinnickinnic River due to the petroleum cleanup efforts. This temporary safety zone is necessary to protect the...-AA00 Safety Zone; Kinnickinnic River Containment and Cleanup; Milwaukee, WI AGENCY: Coast Guard, DHS. ACTION: Temporary final rule. SUMMARY: The Coast Guard is establishing a temporary safety zone on the...

Nirex Safety Assessment Research Programme bibliography, 1987

International Nuclear Information System (INIS)

Cooper, M.J.; Hodgkinson, D.P.

1987-06-01

This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW). All work referred to has been funded, or partly funded, by UK Nirex Limited, previously known as the Nuclear Industry Radioactive Waste Executive (NIREX). The bibliography has been divided into two sections, a List of Publications in roughly chronological order and an Author Index. The topics involved include near-field and far-field studies. The near-field includes the waste, its immobilising medium, its container, the engineered structure in which the containers are emplaced, and the immediately adjacent geological formation disturbed by the construction of the repository. The far-field is the undisturbed geological formation between the near-field and the biosphere. (author)

Regulatory review of safety cases and safety assessments for near surface

International Nuclear Information System (INIS)

Nys, V.

2003-01-01

The activities of the ASAM Regulatory Review Working Group are presented. Regulatory review of the safety assessment is made. It includes the regulatory review of post-closure safety assessment; safety case development and confidence building. The ISAM methodology is reviewed and SA system description is presented. Recommendations on the review process management are given

Safety Auditing and Assessments

Science.gov (United States)

Goodin, James Ronald (Ronnie)

2005-01-01

Safety professionals typically do not engage in audits and independent assessments with the vigor as do our quality brethren. Taking advantage of industry and government experience conducting value added Independent Assessments or Audits benefits a safety program. Most other organizations simply call this process "internal audits." Sources of audit training are presented and compared. A relation of logic between audit techniques and mishap investigation is discussed. An example of an audit process is offered. Shortcomings and pitfalls of auditing are covered.

A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 3: Appendices

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The third volume of the Probabilistic Safety Assessment contains supporting information for the PSA as follows: Appendix C (continued) with details of the system analysis and reports for the system/top event models; Appendix D with results of the specific engineering analyses of internal initiating events; Appendix E, containing supporting data for the human performance assessment,; Appendix F with details of the estimation of the frequency of leaks at HIFAR and Appendix G, containing event sequence model and quantification results

Regulatory review and confidence building in post-closure safety assessments and safety cases for near surface disposal facilities-IAEA ASAM coordinated research programme

International Nuclear Information System (INIS)

Gonzales, A.; Simeonov, G.; Bennett, D.G.; Nys, V.; Ben Belfadhel, M.

2005-01-01

Some years ago, the IAEA successfully concluded a Coordinated Research Program (CRP) called Islam, which focussed on the development of an Improved Safety Assessment Methodology for near-surface radioactive waste disposal facilities. In November 2002, and as an extension of ISAM, the IAEA launched a new CRP called ASAM, designed to test the Application of the Safety Assessment Methodology by considering a range of near-surface disposal facilities. The ASAM work programme is being implemented by three application working groups and two cross-cutting working groups. The application working groups are testing the applicability of the ISAM methodology by assessing an existing disposal facility in Hungary, a copper mine in South Africa, and a hypothetical facility containing heterogenous wastes, such as disused sealed sources. The first cross-cutting working group is addressing a number of technical issues that are common to all near-surface disposal facilities, while the second group, the Regulatory Review Working Group (RRWG) is developing guidance on how to gain confidence in safety assessments and safety cases, and on how to conduct regulatory reviews of safety assessments. This paper provides a brief overview of the work being conducted by the Regulatory Review Working Group. (author)

Confidence building in safety assessments

International Nuclear Information System (INIS)

Grundfelt, Bertil

1999-01-01

Future generations should be adequately protected from damage caused by the present disposal of radioactive waste. This presentation discusses the core of safety and performance assessment: The demonstration and building of confidence that the disposal system meets the safety requirements stipulated by society. The major difficulty is to deal with risks in the very long time perspective of the thousands of years during which the waste is hazardous. Concern about these problems has stimulated the development of the safety assessment discipline. The presentation concentrates on two of the elements of safety assessment: (1) Uncertainty and sensitivity analysis, and (2) validation and review. Uncertainty is associated both with respect to what is the proper conceptual model and with respect to parameter values for a given model. A special kind of uncertainty derives from the variation of a property in space. Geostatistics is one approach to handling spatial variability. The simplest way of doing a sensitivity analysis is to offset the model parameters one by one and observe how the model output changes. The validity of the models and data used to make predictions is central to the credibility of safety assessments for radioactive waste repositories. There are several definitions of model validation. The presentation discusses it as a process and highlights some aspects of validation methodologies

Safety assessment for Area 5 radioactive-waste-management site

International Nuclear Information System (INIS)

Hunter, P.H.; Card, D.H.; Horton, K.

1982-09-01

The Area 5 Radioactive Waste Management Safety Assessment Document contains evaluations of site characteristics, facilities, and operating practices that contribute to the safe handling, storage, and disposal of low-level radioactive wastes at the Nevada Test Site. Physical geography, cultural factors, climate and meteorology, geology, hydrology (with emphasis on radionuclide migration), ecology, natural phenomena, and natural resources are discussed and determined to be suitable for effective containment of radionuclides. A separate section considers facilities and operating practices such as monitoring, storage/disposal criteria, site maintenance, equipment, and support. The section also considers the transportation and waste handling requirements supporting the new Greater Confinement Disposal Facility (GCDF), GCDF demonstration project, and other requirements for the safe handling, storage, and disposal of low-level radioactive wastes. Finally, the document provides an analysis of releases and an assessment of the near-term operational impacts and dose commitments to operating personnel and the general public from normal operations and anticipated accidental occurrences. The conclusion of this report is that the Area 5 Radioactive Waste Management Site is suitable for low-level radioactive waste handling, storage, and disposal. Also, the new GCDF demonstration project will not affect the overall safety of the Area 5 Radioactive Waste Management Site

Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

2014-10-15

The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

Energy Technology Data Exchange (ETDEWEB)

R.M. Dolin

2003-03-03

This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

Determination of Safety Performance Grade of NPP Using Integrated Safety Performance Assessment (ISPA) Program

International Nuclear Information System (INIS)

Chung, Dae Wook

2011-01-01

Since the beginning of 2000, the safety regulation of nuclear power plant (NPP) has been challenged to be conducted more reasonable, effective and efficient way using risk and performance information. In the United States, USNRC established Reactor Oversight Process (ROP) in 2000 for improving the effectiveness of safety regulation of operating NPPs. The main idea of ROP is to classify the NPPs into 5 categories based on the results of safety performance assessment and to conduct graded regulatory programs according to categorization, which might be interpreted as 'Graded Regulation'. However, the classification of safety performance categories is highly comprehensive and sensitive process so that safety performance assessment program should be prepared in integrated, objective and quantitative manner. Furthermore, the results of assessment should characterize and categorize the actual level of safety performance of specific NPP, integrating all the substantial elements for assessing the safety performance. In consideration of particular regulatory environment in Korea, the integrated safety performance assessment (ISPA) program is being under development for the use in the determination of safety performance grade (SPG) of a NPP. The ISPA program consists of 6 individual assessment programs (4 quantitative and 2 qualitative) which cover the overall safety performance of NPP. Some of the assessment programs which are already implemented are used directly or modified for incorporating risk aspects. The others which are not existing regulatory programs are newly developed. Eventually, all the assessment results from individual assessment programs are produced and integrated to determine the safety performance grade of a specific NPP

Containment-emergency-sump performance. Technical findings related to Unresolved Safety Issue A-43

International Nuclear Information System (INIS)

1983-04-01

This report summarizes key technical findings related to the Unresolved Safety Issue A-43, Containment Emergency Sump Performance, and provides recommendations for resolution of attendant safety issues. The key safety questions relate to: (a) effects of insulation debris on sump performance; (b) sump hydraulic performance as determined by design features, submergence, and plant induced effects, and (c) recirculation pump performance wherein air and/or particulate ingestion can occur. The technical findings presented in this report provide information relevant to the design and performance evaluation of the containment emergency sump

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

International Nuclear Information System (INIS)

Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

2008-01-01

Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

A study on safety assessment methodology for a vitrification plant

Energy Technology Data Exchange (ETDEWEB)

Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

2002-03-15

In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

Assessment of the long-term safety for SFR

Energy Technology Data Exchange (ETDEWEB)

Greis Dahlberg, Christina; Vahlund, Frederik [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)

2015-07-01

During operation and decommissioning of the Swedish nuclear facilities, radioactive waste is generated that must be disposed of. Besides waste from the nuclear facilities, some waste derives from other activities such as industry, research, medical care, etc. Short-lived low- and intermediate-level waste from these activities is disposed of in the final repository for short-lived radioactive waste, SFR, in Forsmark. The facility, which has been in operation since 1988, is owned and operated by Svensk Karnbranslehantering AB, SKB. The existing facility has neither sufficient space nor a license to receive decommissioning waste. SFR must therefore be extended so that shortlived low- and intermediate-level decommissioning waste from the nuclear facilities can also be received. The need for additional capacity has been accentuated by the closure of two reactors in Barseback. These reactors cannot be dismantled until the SFR facility has been extended. The existing repository is built to receive, and after closure serve as a passive repository for, low- and intermediate-level radioactive waste. The disposal rooms are situated in the bedrock beneath the sea floor, covered by about 60 metres of rock. The repository has been designed so that it can be abandoned after closure without requiring further measures to maintain its function. The extension of SFR, is done at the -120 m level immediately adjacent to, and within the same depth range as, the existing facility. The basic function of the existing SFR and of the extended one will be the same. However, a clear difference is the design of the tunnel and the rock vault that are required to permit transport and storage of whole reactor pressure vessels. The application for a license to build this extension includes an assessment of the long-term safety (post-closure safety) of the facility. The safety assessment also contains an updated assessment of the long-term safety of the existing facility. The safety assessment for

Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

1983-08-01

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

International Nuclear Information System (INIS)

1992-01-01

The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Independent assessment for new nuclear reactor safety

Directory of Open Access Journals (Sweden)

D'Auria Francesco

2017-01-01

Full Text Available A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs. Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry. The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

Independent assessment for new nuclear reactor safety

International Nuclear Information System (INIS)

D'Auria, F.; Glaeser, H.; Debrecin, N.

2017-01-01

A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty (BEPU) approach. (authors)

Safety assessments using surveillance programmes and data base

International Nuclear Information System (INIS)

Njo, D.H.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 12 presents some aspects on safety assessments of RPV materials during the life of a NPP, using surveillance programmes and data bases. Specific criteria for the usefulness of data bases are developed

Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

Energy Technology Data Exchange (ETDEWEB)

Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-10-26

The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

Development of the safety assessment technology for the radioactive waste disposal

International Nuclear Information System (INIS)

Kim, Chang Lak; Choi, Kwang Sub; Cho, Chan Hee; Lee, Myung Chan; Kim, Jhin Wung

1992-03-01

The major goal of this project is to develop a source-term model for the safety assessment of a low- and intermediate-level radioactive waste repository as follows: 1) estimation of the arising of low- and intermediate-level radioactive wastes, 2) development of inventory data base, 3) development of a source-term code for shallow-land disposal, and 4) improvement of the REPS source-term code for rock cavern type disposal developed already in 1990 and conservative safety assessment for an imaginary repository. In addition, the source of C-14 in the inventory is assessed by two methods: decontamination factor and scaling factor. The source-term code for shallow-land disposal include the following submodels: surface water penetration into the repository, concrete degradation, corrosion of container drums, leaching of radionuclides from waste forms, and migration of radionuclides from engineered disposal facility is estimated by this code. (Author)

Structural capacity assessment of a generic pre-stressed concrete containment structure under aircraft impact

International Nuclear Information System (INIS)

Iliev, Alexander

2013-01-01

The studied containment expressed adequate capacity to resist impact loads in the upper range of the studied diapason. The aircraft impact capacity of the containment for impact in the upper part of the cylindrical shell is about 25‐30% higher than the capacity for impact in the middle part of the cylindrical shell. The obtained fragility curves reefed to MoA can be then used for various additional calculations in the safety assessment of nuclear facilities under aircraft impact

Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

International Nuclear Information System (INIS)

Briggs, J.B.

1991-02-01

Intraplant shipment of small quantities of plutonium and uranium at the Rocky Flats Plant (RFP) are made in one-gallon shipping containers. Criticality safety calculations have been performed to provide an analytical basis upon which handling, storage, and transportation limits on these containers are based. The calculations and results are documented in this report. This analysis was categorized as Quality Level A (according to the EG ampersand G Idaho Quality Manual) in that it is a service whose failure could cause undue risks to employees or public health and safety. It is intended to comply with NQA-1. 7 refs., 7 figs., 12 tabs

The in-depth safety assessment (ISA) pilot projects in Ukraine

International Nuclear Information System (INIS)

Kot, C. A.

1998-01-01

Ukraine operates pressurized water reactors of the Soviet-designed type, VVER. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs). After approval of the SARS by the Ukrainian Nuclear Regulatory Authority, the plants will be granted longer-term operating licenses. In September 1995, the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine issued a new contents requirement for the safety analysis reports of VVERs in Ukraine. It contains requirements in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The DBA requirements are an expanded version of the older SAR requirements. The last two requirements, on PRA and BDBA, are new. The US Department of Energy (USDOE), through the International Nuclear Safety Program (INSP), has initiated an assistance and technology transfer program to Ukraine to assist their nuclear power stations in developing a Western-type technical basis for the new SARS. USDOE sponsored In-Depth Safety Assessments (ISAs) have been initiated at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1. USDOE/INSP have structured the ISA program in such a way as to provide maximum assistance and technology transfer to Ukraine while encouraging and supporting the Ukrainian plants to take the responsibility and initiative and to perform the required assessments

AGR core safety assessment methodologies

International Nuclear Information System (INIS)

McLachlan, N.; Reed, J.; Metcalfe, M.P.

1996-01-01

To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

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

International Nuclear Information System (INIS)

2010-01-01

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

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

International Nuclear Information System (INIS)

Ohi, Takao; Kawasaki, Daisuke; Chiba, Tamotsu; Takase, Toshio; Hane, Koji

2013-01-01

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner. (author)

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan [and others

2004-08-01

This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8

Interim main report of the safety assessment SR-Can

International Nuclear Information System (INIS)

Hedin, Allan

2004-08-01

This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10 -6 for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8. Hydrogeological

A Probabilistic Methodology for Assessing the Effectiveness of the Containment Filtered Venting Systems

Energy Technology Data Exchange (ETDEWEB)

Kim, Namyeong; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2015-10-15

After the Chernobyl nuclear accident, mainly in Sweden, Germany, France, Switzerland, the Netherlands and other European countries have installed CFVS. In the US, some Boiling Water Reactor type only the voluntary installation of CFVS was required. But until now it has not been installed for pressurized water reactors. In Korea, CFVS is currently installed on Wolseong Unit 1 and preferentially applied to Heavy Water Reactor. Later it plans to apply for the Light Water Reactor. In this study, a safety improvement of installing the CFVS was assessed by the tool of Probabilistic Safety Assessment (PSA) for a reference plant. The CFVS is under installment in CANDU reactor for preventing the containment failure during severe accidents. But it has been evaluated that the effectiveness is negligible because of adverse effects of radioactive nuclides releases. Now the CFVS has not been installed yet in the LWR. The results can vary greatly depending on the detailed assessment.

Procedures for self-assessment of operational safety

International Nuclear Information System (INIS)

1997-08-01

Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and and the safety culture as a whole. The concepts developed in this report present the basic approach to self-assessment taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject

Safety analysis report for packaging (onsite) decontaminated equipment self-container

International Nuclear Information System (INIS)

Boehnke, W.M.

1998-01-01

The purpose of this Safety Analysis Report for Packaging (SARP) is to demonstrate that specific decontaminated equipment can be safely used as its own self-container. As a Decontaminated Equipment Self-Container (also referred to as a self-container), no other packaging, such as a burial box, would be required to transport the equipment onsite. The self-container will consist of a piece of equipment or apparatus which has all readily removable interior contamination removed, all of its external openings sealed, and all external surfaces decontaminated to less than 2000 dpm/100 cm for gamma-emitting radionuclides and less than 220 dpm/100 CM2 for alpha-emitting radionuclides

Assessment of occupational exposure due to intakes of radionuclides. Safety guide

International Nuclear Information System (INIS)

2004-01-01

Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

Assessment of occupational exposure due to intakes of radionuclides. Safety guide

International Nuclear Information System (INIS)

2000-01-01

Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

Assessment of occupational exposure due to intakes of radionuclides. Safety guide

International Nuclear Information System (INIS)

2001-01-01

Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

Assessment of occupational exposure due to intakes of radionuclides. Safety guide

International Nuclear Information System (INIS)

2002-01-01

Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

Assessment of occupational exposure due to intakes of radionuclides. Safety guide

International Nuclear Information System (INIS)

1999-01-01

Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

Uncertainty analysis in safety assessment

International Nuclear Information System (INIS)

Lemos, Francisco Luiz de; Sullivan, Terry

1997-01-01

Nuclear waste disposal is a very complex subject which requires the study of many different fields of science, like hydro geology, meteorology, geochemistry, etc. In addition, the waste disposal facilities are designed to last for a very long period of time. Both of these conditions make safety assessment projections filled with uncertainty. This paper addresses approaches for treatment of uncertainties in the safety assessment modeling due to the variability of data and some current approaches used to deal with this problem. (author)

NUMO's approach for long-term safety assessment - 59404

International Nuclear Information System (INIS)

Ebashi, Takeshi; Kaku, Kenichi; Ishiguro, Katsuhiko

2012-01-01

One of NUMO's policies for ensuring safety is staged and flexible project implementation and decision-making based on iterative confirmation of safety. The safety assessment takes the central role in multiple lines of reasoning and argumentation by providing a quantitative evaluation of long-term safety; a key aspect is uncertainty management. This paper presents NUMO's basic strategies for long-term safety assessment based on the above policy. NUMO's approach considering Japanese boundary conditions is demonstrated as a starting-point for evaluating the long-term safety of an actual site. In Japan, the Act on Final Disposal of Specified Radioactive Waste states that the siting process shall consist of three stages. The Nuclear Waste Management Organization of Japan (NUMO) is responsible for geological disposal of vitrified high-level waste and some types of TRU waste. NUMO has chosen to implement a volunteer approach to siting. NUMO decided to prepare the so-called 2010 technical report, which sets out three safety policies, one of which is staged project implementation and decision-making based on iterative confirmation of safety. Based on this policy, NUMO will gradually integrate relevant interdisciplinary knowledge to build a safety case when a formal volunteer application is received that would allow site investigations to be initiated. The safety assessment takes the central role in multiple lines of reasoning and argumentation by providing a quantitative evaluation of long-term safety; one of a key aspect is uncertainty management. This paper presents the basic strategies for NUMO's long-term safety assessment based on the above policy. In concrete terms, the common procedures involved in safety assessment are applied in a stepwise manner, based on integration of knowledge obtained from site investigations/evaluations and engineered measures. The results of the safety assessment are then reflected in the planning of site investigations and engineered

Probabilistic safety assessment

International Nuclear Information System (INIS)

Hoertner, H.; Schuetz, B.

1982-09-01

For the purpose of assessing applicability and informativeness on risk-analysis methods in licencing procedures under atomic law, the choice of instruments for probabilistic analysis, the problems in and experience gained in their application, and the discussion of safety goals with respect to such instruments are of paramount significance. Naturally, such a complex field can only be dealt with step by step, making contribution relative to specific problems. The report on hand shows the essentials of a 'stocktaking' of systems relability studies in the licencing procedure under atomic law and of an American report (NUREG-0739) on 'Quantitative Safety Goals'. (orig.) [de

Screening-Level Safety Assessment of Personal Care Product Constituents Using Publicly Available Data

Directory of Open Access Journals (Sweden)

Ernest S. Fung

2018-06-01

Full Text Available Organizations recommend evaluating individual ingredients when assessing the safety of personal care or cosmetic products. The goal of this study was to present a screening-level safety assessment methodology to evaluate the safety of a product by identifying individual ingredients, determining their frequency of use in on-market products, and examining published safe-level-of-use information for each ingredient. As a case study, we evaluated WEN by Chaz Dean (WCD cleansing conditioners since there have been claims of adverse health effects associated with product use. We evaluated 30 ingredients in three on-market WCD cleansing conditioners. We then analyzed the National Library of Medicine’s Household Products Database and the Environmental Working Group’s (EWG Skin Deep Cosmetic Database, two of the largest publicly available databases, for other on-market personal care and cosmetic products that contained these ingredients. Safe-level-of-use information for each ingredient was obtained by reviewing peer-reviewed literature, the Food and Drug Administration’s (FDA generally recognized as safe (GRAS database, available Cosmetic Ingredient Review (CIR publications, and available product safety publications. The results of this analysis showed that more than 20,000 personal care and cosmetic products contained one or more of the evaluated ingredients used in WCD cleaning conditioners. Published safety information was available for 21 of the 30 evaluated ingredients: seven identified ingredients were designated as GRAS by the FDA and 16 ingredients had safe-level-of-use information available from the CIR. This study presents a screening-level safety assessment methodology that can serve as an initial screening tool to evaluate the safety of an ingredient intended for use in personal care and cosmetic products before a product is launched onto the market. This study provides evidence that the evaluated WCD cleansing conditioner ingredients

Life cycle assessment of façade coating systems containing manufactured nanomaterials

Energy Technology Data Exchange (ETDEWEB)

Hischier, Roland, E-mail: roland.hischier@empa.ch; Nowack, Bernd [Swiss Federal Laboratories for Materials Science and Technology (Empa), Technology and Society Lab (TSL) (Switzerland); Gottschalk, Fadri [ETSS (Switzerland); Hincapie, Ingrid [Swiss Federal Laboratories for Materials Science and Technology (Empa), Technology and Society Lab (TSL) (Switzerland); Steinfeldt, Michael [University of Bremen FB 4/FG 10 Technological Design and Development (Germany); Som, Claudia [Swiss Federal Laboratories for Materials Science and Technology (Empa), Technology and Society Lab (TSL) (Switzerland)

2015-02-15

Nanotechnologies are expected to hold considerable potential for the development of new materials in the construction sector. Up to now the environmental benefits and risks of products containing manufactured nanomaterials (MNM) have been quantified only to a limited extent. This study aims to assess the potential environmental, health and safety impacts of coatings containing MNM using Life-cycle assessment: Do paints containing MNM result in a better environmental performance than paints not containing MNM? The study shows that the results depend on a number of factors: (i) The MNM have to substitute an (active) ingredient of the initial paint composition and not simply be an additional ingredient. (ii) The new composition has to extend the lifetime of the paint for such a time period that the consumption of paint along the life cycle of a building is reduced. (iii) Releases of MNM have to be reduced to the lowest level possible (in particular by dumping unused paint together with the packaging). Only when all these boundary conditions are fulfilled, which is the case only for one of the three paint systems examined, is an improved environmental performance of the MNM-containing paint possible for the paint compositions examined in this study.

A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

Energy Technology Data Exchange (ETDEWEB)

Park, Sunghyun; Kim, Namyeong; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2015-10-15

In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant.

A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Park, Sunghyun; Kim, Namyeong; Jae, Moosung

2015-01-01

In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant

Assessment of basic safety issues

International Nuclear Information System (INIS)

Queniart, D.

1996-01-01

Work on the French-German common safety approach for future nuclear power plants continued in 1994 to allow for more detailed discussion of some major issues, taking into account the options provided by the industry for the EPR (European Pressurized water Reactor) project, as described in the document entitled 'Conceptual Safety Features Review File'. Seven meetings of a GPR/RSK advisory experts subgroup, six GPR/RSK plenary sessions and six meetings of the safety authorities (DFD) dealt with the following topics: design of the systems and use of probabilistic approaches, application of a 'break preclusion' approach to the main primary pipings, protection against external hazards (aircraft crashes, explosions, earthquakes), provisions with respect to accidents involving core melt and to containment design, radiological consequences of reference accidents and accidents involving core melt at low pressure. The important aspects of the joint policy are recalled in the presentation. The whole set of GPR/RSK recommendations were agreed by the French and German safety authorities during the DFD meetings of 1994 and early 1995. The utilities decided to begin the basic design phase in February, 1995. Work is now continuing to develop the common French-German approach for future nuclear power plants, in the same way as before. In 1995, this mainly covers the design of the containment and of the systems, but also new issues such as the protection against secondary side overpressurization, radiological protection of workers and radioactive wastes. (J.S.). 3 figs., 1 tab

Complementary safety assessments - Report by the French Nuclear Safety Authority

International Nuclear Information System (INIS)

2011-12-01

As an immediate consequence of the Fukushima accident, the French Authority of Nuclear Safety (ASN) launched a campaign of on-site inspections and asked operators (mainly EDF, AREVA and CEA) to make complementary assessments of the safety of the nuclear facilities they manage. The approach defined by ASN for the complementary safety assessments (CSA) is to study the behaviour of nuclear facilities in severe accidents situations caused by an off-site natural hazard according to accident scenarios exceeding the current baseline safety requirements. This approach can be broken into 2 phases: first conformity to current design and secondly an approach to the beyond design-basis scenarios built around the principle of defence in depth. 38 inspections were performed on issues linked to the causes of the Fukushima crisis. It appears that some sites have to reinforce the robustness of the heat sink. The CSA confirmed that the processes put into place at EDF to detect non-conformities were satisfactory. The complementary safety assessments demonstrated that the current seismic margins on the EDF nuclear reactors are satisfactory. With regard to flooding, the complementary safety assessments show that the complete reassessment carried out following the flooding of the Le Blayais nuclear power plant in 1999 offers the installations a high level of protection against the risk of flooding. Concerning the loss of electrical power supplies and the loss of cooling systems, the analysis of EDF's CSA reports showed that certain heat sink and electrical power supply loss scenarios can, if nothing is done, lead to core melt in just a few hours in the most unfavourable circumstances. As for nuclear facilities that are not power or experimental reactors, some difficulties have appeared to implement the CSA approach that was initially devised for reactors. Generally speaking, ASN considers that the safety of nuclear facilities must be made more robust to improbable risks which are not

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Safety assessment methodologies for near surface disposal facilities. Results of a co-ordinated research project (ISAM). Volume 1: Review and enhancement of safety assessment approaches and tools. Volume 2: Test cases

International Nuclear Information System (INIS)

2004-07-01

the Safety Guide on 'Safety Assessment for Near Surface Disposal of Radioactive Waste' (Safety Standards Series No. WS-G- 1.1). The report of this CRP is presented in two volumes; Volume 1 contains a summary and a complete description of the ISAM project methodology and Volume 2 presents the application of the methodology to three hypothetical test cases

Safety culture' is integrating 'human' into risk assessment

International Nuclear Information System (INIS)

Sugimoto, Taiji

2014-01-01

Significance of Fukushima nuclear power accident requested reconsideration of safety standards, of which we had usually no doubt. Risk assessment standard (JIS B 9702), Which was used for repetition of database preparation and cumulative assessment, defined allowable risk and residual risk. However, work site and immediate assessment was indispensable beside such assessment so as to ensure safety. Risk of casualties was absolutely not acceptable in principle and judgments to approve allowable risk needed accountability, which was reminded by safety culture proposed by IAEA and also identified by investigation of organizational cause of Columbia accident. Actor of safety culture would be organization and individual, and mainly individual. Realization of safety culture was conducted by personnel having moral consciousness and firm sense of mission in the course of jobs and working daily with sweat pouring. Safety engineering/technology should have framework integrating human as such totality. (T. Tanaka)

Nirex safety assessment research programme: annual report for 1985/86

International Nuclear Information System (INIS)

Hodgkinson, D.P.; Cooper, M.J.

1987-01-01

The purpose of this report is to provide information for post-emplacement radiological safety assessment relating to the disposal of intermediate-level and low-level radioactive wastes into underground repositories and the sea bed. Topics reported are chemical equilibrium studies, laser spectroscopy, the corrosion of containers, properties of concretes, microbiology, transport in clays, the behaviour of sea disposal packages and leaching from cements in seawater. There is close contact between experimental work and mathematical modelling. (U.K.)

ITER safety

International Nuclear Information System (INIS)

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

1991-01-01

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

Novi Han Radioactive Waste Repository post-closure safety assessment, ver.2

International Nuclear Information System (INIS)

Mateeva, M.

2003-01-01

The methodology for the post-closure safety assessment is presented. The assessment context includes regulatory framework (protection principles); scope and time frame; radiological and technical requirements; modeling etc. The description of the Novi Han disposal system contains site location. meteorological, hydrological and seismological characteristics; waste and repository description and human activities characteristics. The next step in the methodology is scenario development and justification. The systematic generation os exposure scenarios is considered as central to the post-closure safety assessment. The most important requirements for the systematic scenario generation approach are: transparency, comprehensiveness (all possible FEPs influencing the the disposal system and the radionuclide release should be considered); relevant future evolutions; identification of critical issues and investigation of the robustness of the system. For the source-pathway-receptor analysis the Process System is divided into near-field, geosphere/atmosphere and biosphere, describing the key facets controlling the potential radionuclide migration to the environment. The schematic division of the Novi Han near-field Process System into lower-level conceptual features is presented and discussed. As a result of the examinations of the FEPs three classes of scenarios are identified for the Novi Han post-closure safety assessment: Environmental evolution scenarios (geological change and climate change); future human action scenarios (human intrusion and archaeological action); Scenarios with very low probability (terrorism, crashes, explosions). The safety assessment iteration leads to identification of a modern scenario generation approach, assessment of key radionuclide releases, geological and hydrological evaluation, identification of the key parameters from sensitivity analysis etc. Examples of conceptual models are given. For the mathematical modeling the AMBER code is used

Safety of systems for the retention of wastes containing radionuclides

International Nuclear Information System (INIS)

1980-11-01

Information and minimal requirements demanded by CNEN for the emission of the Approval Certificate of the Safety Analysis Report related to system for the retention of wastes containing radionuclide, are established, aiming to assure low radioactivity levels to the environment. (E.G.) [pt

FEP report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This report documents the analysis and processing of features, events and processes, FEPs, that has been carried out within the safety assessment SR-Site, and forms an important part of the reporting of the project. The main part of the work was conducted within the earlier safety assessment SR-Can, which was a preparatory stage for the SR-Site assessment. The overall objective of the FEP analysis and processing in both SR-Can and SR-Site included development of a database of features, events and processes, an SKB FEP database, in a format that facilitates both a systematic analysis of FEPs and documentation of that FEP analysis, as well as facilitating revisions and updates to be made in connection with new safety assessments. The primary objective in SR-Site was to establish an SR-Site FEP catalogue within the framework of the SKB FEP database. This FEP catalogue was required to contain all FEPs that needed to be handled in SR-Site and is an update of the corresponding SR-Can FEP catalogue that was established for the SR-Can assessment. The starting point for the handling of FEPs in SR-Site was the SR-Can version of the SKB FEP database and associated SR-Can reports. The SR-Can version of the SKB FEP database includes the SR-Can FEP catalogue, as well as the sources for the identification of FEPs in SR-Can, namely the SR 97 processes and variables, Project FEPs in the NEA International FEP database version 1.2 and matrix interactions in the Interaction matrices developed for a deep repository of the KBS-3 type. Since the completion of the FEP work within SR-Can, an updated electronic version, version 2.1, of the NEA FEP database has become available. Compared with version 1.2 of the NEA FEP database, version 2.1 contains FEPs from two more projects. As part of SR-Site, all new Project FEPs in version 2.1 of the NEA FEP database have been mapped according to the methodology adopted in SR-Can resulting in an SR-Site version of the SKB FEP database. The SKB FEP

FEP report for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

This report documents the analysis and processing of features, events and processes, FEPs, that has been carried out within the safety assessment SR-Site, and forms an important part of the reporting of the project. The main part of the work was conducted within the earlier safety assessment SR-Can, which was a preparatory stage for the SR-Site assessment. The overall objective of the FEP analysis and processing in both SR-Can and SR-Site included development of a database of features, events and processes, an SKB FEP database, in a format that facilitates both a systematic analysis of FEPs and documentation of that FEP analysis, as well as facilitating revisions and updates to be made in connection with new safety assessments. The primary objective in SR-Site was to establish an SR-Site FEP catalogue within the framework of the SKB FEP database. This FEP catalogue was required to contain all FEPs that needed to be handled in SR-Site and is an update of the corresponding SR-Can FEP catalogue that was established for the SR-Can assessment. The starting point for the handling of FEPs in SR-Site was the SR-Can version of the SKB FEP database and associated SR-Can reports. The SR-Can version of the SKB FEP database includes the SR-Can FEP catalogue, as well as the sources for the identification of FEPs in SR-Can, namely the SR 97 processes and variables, Project FEPs in the NEA International FEP database version 1.2 and matrix interactions in the Interaction matrices developed for a deep repository of the KBS-3 type. Since the completion of the FEP work within SR-Can, an updated electronic version, version 2.1, of the NEA FEP database has become available. Compared with version 1.2 of the NEA FEP database, version 2.1 contains FEPs from two more projects. As part of SR-Site, all new Project FEPs in version 2.1 of the NEA FEP database have been mapped according to the methodology adopted in SR-Can resulting in an SR-Site version of the SKB FEP database. The SKB FEP

Uncertainty analysis in safety assessment

Energy Technology Data Exchange (ETDEWEB)

Lemos, Francisco Luiz de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Sullivan, Terry [Brookhaven National Lab., Upton, NY (United States)

1997-12-31

Nuclear waste disposal is a very complex subject which requires the study of many different fields of science, like hydro geology, meteorology, geochemistry, etc. In addition, the waste disposal facilities are designed to last for a very long period of time. Both of these conditions make safety assessment projections filled with uncertainty. This paper addresses approaches for treatment of uncertainties in the safety assessment modeling due to the variability of data and some current approaches used to deal with this problem. (author) 13 refs.; e-mail: lemos at bnl.gov; sulliva1 at bnl.gov

The Design of Cooling System Model on The AP1000 Containment

International Nuclear Information System (INIS)

Daddy Setyawan; Yerri Noer Kartiko; Aryadi Suwono; Ari Darmawan Pasek; Nathanael P Tandian; Efrizon Umar

2009-01-01

The policy of national energy leads to the utilization of new energy as nuclear energy, and also contains some efforts to increase reactor safety and optimizing in the design of safety system component such as passive cooling system on reactor containment tank. Because of this, the assessment of safety level to passive safety system needs to be made. To increase the understanding it, the design of cooling system model on containment tank should be done to get safety level on cooling system in the AP1000 containment. To reach the similar model with reality and inexpensive cost, we should make assessment about similarity and dimensionless number. While the heat transfer of air natural circulation and water spray cooling system are a result of gravity approach, we can calculate Grashof modification number and Reynolds number respectively. By this approach, we have a factor of forty for laboratory model. From this model, we hope that we get characteristic correlation to heat transfer on the containment of AP1000 for both air natural circulation and water spray result from gravity. Finally, we can assess the safety level of passive cooling system on the AP1000 containment. (author)

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan (ed.) [and others

2004-08-01

This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection

Safety aspects of NPP ageing

International Nuclear Information System (INIS)

1995-01-01

Preparation of safety practices on assessment and management of aging of major NPP components important to safety, CRP on management of aging of concrete containment buildings, CRP on management of aging of in-containment instrumentation and control cables are outlined

The Nordic safety program on accident consequence assessment

International Nuclear Information System (INIS)

Tveten, U.

1988-01-01

One important part of Nordic cooperation is partially funded by the Nordic Council of Ministers, namely the work performed within the Nordic Safety Program (often referred to as the NKA projects). NKA is the Nordic abbreviation of the Nordic Liaison Committee on Atomic Energy. One program area in the present four-year period is concerned with problems related to reactor accident consequence assessment, and contains almost twenty projects covering a wide range of subjects. The author is program coordinator for this program area. The program will be completed in 1989. The program was strongly influenced by Chernobyl, and a number of new projects were included in the program in 1986. Involved in the program are these Nordic institutions: Riso National Laboratory (Denmark). Technical Research Centre of Finland. Finnish Centre for Radiation and Nuclear Safety. Finnish Meteorological Institute. Institute for Energy Technology (Norway). Agricultural University of Norway. Meteorological Institute of Norway. Studsvik Energiteknik AB (Sweden). National Defence Research Laboratory (Sweden)

Safety assessment of foods derived from genetically modified crops

NARCIS (Netherlands)

Kleter, G.A.; Kuiper, H.A.

2003-01-01

The pre-market safety assessment of foods derived from genetically modified crops is carried out according to the consensus approach of "substantial equivalence", in other words: the comparative safety assessment. Currently, the safety assessment of genetically modified foods is harmonized at the

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

International Nuclear Information System (INIS)

2003-01-01

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

General safety considerations

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

General safety considerations

International Nuclear Information System (INIS)

2001-01-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

General safety considerations

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-09-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

General safety considerations

International Nuclear Information System (INIS)

1998-01-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

Assessment of safety culture: Changing regulatory approach in Hungary

International Nuclear Information System (INIS)

Ronaky, Jozsef; Toth, Andras

2002-01-01

Hungarian Atomic Energy Authority (HAEA) is changing its inspection practice and assessment methods of safety performance and safety culture in operating nuclear facilities. The new approach emphasises integrated team inspection of safety cornerstones and systematic assessment of safety performance of operators. (author)

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

Directory of Open Access Journals (Sweden)

Costica Bejinariu

2017-07-01

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

Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

International Nuclear Information System (INIS)

Ewing, D.J.F.; Campbell, J.F.

1994-01-01

This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

The radiation safety self-assessment program of Ontario Hydro

International Nuclear Information System (INIS)

Armitage, G.; Chase, W.J.

1987-01-01

Ontario Hydro has developed a self-assessment program to ensure that high quality in its radiation safety program is maintained. The self-assessment program has three major components: routine ongoing assessment, accident/incident investigation, and detailed assessments of particular radiation safety subsystems or of the total radiation safety program. The operation of each of these components is described

Probabilistic safety assessment for seismic events

International Nuclear Information System (INIS)

1993-10-01

This Technical Document on Probabilistic Safety Assessment for Seismic Events is mainly associated with the Safety Practice on Treatment of External Hazards in PSA and discusses in detail one specific external hazard, i.e. earthquakes

Safety analysis and risk assessment handbook

International Nuclear Information System (INIS)

Peterson, V.L.; Colwell, R.G.; Dickey, R.L.

1997-01-01

This Safety Analysis and Risk Assessment Handbook (SARAH) provides guidance to the safety analyst at the Rocky Flats Environmental Technology Site (RFETS) in the preparation of safety analyses and risk assessments. Although the older guidance (the Rocky Flats Risk Assessment Guide) continues to be used for updating the Final Safety Analysis Reports developed in the mid-1980s, this new guidance is used with all new authorization basis documents. With the mission change at RFETS came the need to establish new authorization basis documents for its facilities, whose functions had changed. The methodology and databases for performing the evaluations that support the new authorization basis documents had to be standardized, to avoid the use of different approaches and/or databases for similar accidents in different facilities. This handbook presents this new standardized approach. The handbook begins with a discussion of the requirements of the different types of authorization basis documents and how to choose the one appropriate for the facility to be evaluated. It then walks the analyst through the process of identifying all the potential hazards in the facility, classifying them, and choosing the ones that need to be analyzed further. It then discusses the methods for evaluating accident initiation and progression and covers the basic steps in a safety analysis, including consequence and frequency binning and risk ranking. The handbook lays out standardized approaches for determining the source terms of the various accidents (including airborne release fractions, leakpath factors, etc.), the atmospheric dispersion factors appropriate for Rocky Flats, and the methods for radiological and chemical consequence assessments. The radiological assessments use a radiological open-quotes templateclose quotes, a spreadsheet that incorporates the standard values of parameters, whereas the chemical assessments use the standard codes ARCHIE and ALOHA

Living probabilistic safety assessment (LPSA)

International Nuclear Information System (INIS)

1999-08-01

Over the past few years many nuclear power plant organizations have performed probabilistic safety assessments (PSAs) to identify and understand key plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. PSA is an effective tool for this purpose as it assists plant management to target resources where the largest benefit to plant safety can be obtained. However, any PSA which is to be used in this way must have a credible and defensible basis. Thus, it is very important to have a high quality 'living PSA' accepted by the plant and the regulator. With this background in mind, the IAEA has prepared this report on Living Probabilistic Safety Assessment (LPSA) which addresses the updating, documentation, quality assurance, and management and organizational requirements for LPSA. Deficiencies in the areas addressed in this report would seriously reduce the adequacy of the LPSA as a tool to support decision making at NPPs. This report was reviewed by a working group during a Technical Committee Meeting on PSA Applications to Improve NPP Safety held in Madrid, Spain, from 23 to 27 February 1998

Probabilistic safety assessment in nuclear power plant management

International Nuclear Information System (INIS)

Holloway, N.J.

1989-06-01

Probabilistic Safety Assessment (PSA) techniques have been widely used over the past few years to assist in understanding how engineered systems respond to abnormal conditions, particularly during a severe accident. The use of PSAs in the design and operation of such systems thus contributes to the safety of nuclear power plants. Probabilistic safety assessments can be maintained to provide a continuous up-to-date assessment (Living PSA), supporting the management of plant operations and modifications

Risk assessment on the use of herbal medicinal products containing pyrrolizidine alkaloids.

Science.gov (United States)

Allgaier, Clemens; Franz, Stephanie

2015-11-01

Pyrrolizidine alkaloids (PA) are common plantal toxins directed against insect herbivores. Unsaturated PAs are known to be hepatotoxic. Many of the PAs are in addition mutagenic and some may possibly be carcinogenic for humans. The risk of an exposure to PAs associated with their occurrence in herbal medicinal products and in foodstuff is under current discussion. The present risk assessment for herbal medicinal products containing PAs is based on a margin of safety derivation for foodstuff indicating that a life-long exposure to maximally 0.007 μg/kg bw/day is not expected to be associated with safety concerns. This approach offers a possibility to estimate the potential risk of PA-containing herbal medicinal products irrespective of the route of administration. It assumes PA levels in the final herbal medicinal product below 0.01 ppm and considers for dermal administration a 100% skin penetration of the PAs reflecting a worst-case scenario. As a result, the calculated margins of safety show a potential exposure using herbal medicinal products 70-, 45.5-, and 19.3-fold lower on a one-day base and 608-, 396-, and 168- fold lower on a one-year base for adults, children aged 12 years, and children aged 4 years, respectively, than the thresholds considered acceptable for foodstuff. Copyright © 2015 Elsevier Inc. All rights reserved.

Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

International Nuclear Information System (INIS)

Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

In vitro and in vivo evaluation of efficacy and safety of photoprotective formulations containing antioxidant extracts

Directory of Open Access Journals (Sweden)

Maria Cristina P.P. Reis Mansur

Full Text Available ABSTRACT Chronic exposure to solar radiation could contribute to premature skin aging and skin cancer. Skin presents its own antioxidant defense, however when defenses are out of balance, reactive oxygen species could damage biological structures. In the present work, an oil-in-water photoprotective emulsion was developed and Bauhinia microstachya var. massambabensis Vaz, Fabaceae, extracts at 1% (obtained by extraction with different solvents were added to this emulsion. In vitro and in vivo efficacy and safety of the formulations were evaluated. Spectrophotometric methods and in vivo Colipa test were performed to evaluated efficacy of the formulations, through sun protection factor (SPF determination and UVA protection factor assessment. To the in vitro safety assessment HET-CAM, CAM-TBS and Red Blood Cell tests were performed. Results showed that both extracts contributed to a higher in vivo photoprotection (SPF 18 when compared to the formulation without extract (SPF 13, this result could be attributed to the antioxidant activity of the plant extracts that act by capturing reactive oxygen species. Concerning safety, all formulations were considered non-irritant according to in vitro tests. Formulations containing extracts could be considered efficient and safe for cosmetic use since they presented higher sun protection factor and passed the toxicity tests.

Safety Assessment of Probiotics

Science.gov (United States)

Lahtinen, Sampo J.; Boyle, Robert J.; Margolles, Abelardo; Frias, Rafael; Gueimonde, Miguel

Viable microbes have been a natural part of human diet throughout the history of mankind. Today, different fermented foods and other foods containing live microbes are consumed around the world, including industrialized countries, where the diet has become increasingly sterile during the last decades. By definition, probiotics are viable microbes with documented beneficial effects on host health. Probiotics have an excellent safety record, both in humans and in animals. Despite the wide and continuously increasing consumption of probiotics, adverse events related to probiotic use are extremely rare. Many popular probiotic strains such as lactobacilli and bifidobacteria can be considered as components of normal healthy intestinal microbiota, and thus are not thought to pose a risk for the host health - in contrast, beneficial effects on health are commonly reported. Nevertheless, the safety of probiotics is an important issue, in particular in the case of new potential probiotics which do not have a long history of safe use, and of probiotics belonging to species for which general assumption of safety cannot be made. Furthermore, safety of probiotics in high-risk populations such as critically ill patients and immunocompromized subjects deserves particular attention, as virtually all reported cases of bacteremia and fungemia associated with probiotic use, involve subjects with underlying diseases, compromised immune system or compromised intestinal integrity.

Types of safety assessments of near surface repository for radioactive waste

International Nuclear Information System (INIS)

Mateeva, M.

2004-01-01

The purpose of this article is to presents the classification of different types safety assessments of near surface repository for low and intermediate level radioactive waste substantiated with results of safety assessments generated in Bulgaria. The different approach of safety assessments applied for old existing repository as well as for site selection for construction new repository is outlined. The regulatory requirements in Bulgaria define three main types of assessments: Safety assessment; Technical substation of repository safety; Assessment of repository influence on environment that is in form of report prepared from the Ministry of environment and waters on the base of results obtained in two first types of assessments. Additionally first type is subdivided in three categories - preliminary safety assessment, safety assessment and post closure safety assessment, which are generated using deterministic approach. The technical substation of repository safety is generated using probabilistic approach. Safety assessment results that are presented here are based on evaluation of existing old repository type 'Radon' in Novi Han and real site selection procedure for new near surface repository for low and intermediate level radioactive waste from nuclear power station in Kozloduy. The important role of safety assessment for improvement the repository safety as well as for repository licensing, correct site selection and right choice of engineer barriers and repository design is discussed using generated results. (author)

Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

International Nuclear Information System (INIS)

Shaw, M.E.

1991-12-01

Criticality safety calculations have been performed to provide an analytical basis for handling, storage and transport of Rocky Flats Plant (RFP) one-gallon shipping containers. A mass limit was establish for metal (solid uranium or plutonium) and slurries (undissolved U or Pu solids in a ''mud,'' ''sludge,'' or ''slurry''). A separate volume limit was developed for plutonium solutions (liquids, either aqueous or organic, containing no visible undissolved solids)

Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

Energy Technology Data Exchange (ETDEWEB)

Shaw, M.E.

1991-12-01

Criticality safety calculations have been performed to provide an analytical basis for handling, storage and transport of Rocky Flats Plant (RFP) one-gallon shipping containers. A mass limit was establish for metal (solid uranium or plutonium) and slurries (undissolved U or Pu solids in a ``mud,`` ``sludge,`` or ``slurry``). A separate volume limit was developed for plutonium solutions (liquids, either aqueous or organic, containing no visible undissolved solids).

Development of a numerical tool for safety assessment and emergency management of experimental reactors

International Nuclear Information System (INIS)

Maas, L.; Beuter, A.; Seropian, C.

2010-01-01

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. Among its duties, one important item is to provide help for emergency situations management in case of an accident occurring in a French nuclear facility. In this framework, IRSN develops and applies numerical tools dealing with containment management issues. Up to now IRSN has not got any specific tool for experimental reactors. Accordingly, it has been then decided to extend the ASTEC code, devoted to severe accident scenarios for Pressurized Water Reactors, to this kind of reactors. This lumped-parameter code, co-developed by IRSN and GRS (Germany), covers the entire phenomenology from the initiating event up to fission products release outside the reactor containment, except for the steam explosion and the mechanical integrity of the containment. A first application to experimental reactors was carried out to assess the High Flux Reactor (HFR) operator's improvement proposal concerning the containment management during accidental situations. This reactor, located in Grenoble (France), is composed of a double wall containment with a pressurized containment annulus preventing any direct leakage into the environment. Until now, in case of severe accidents (mainly core melting in pool, explosive reactivity accident called BORAX), the HFR emergency management consisted in isolating the containment building in the early stage of the accident, to prevent any radioactive products release to the environment. The operator decided to improve this containment management during accidental situations by using an air filtering venting system able to maintain a slight sub-atmospheric pressure in the reactor building. The operator's demonstration of the efficiency of this new system is mainly based on containment pressure evaluations during accidental transients. IRSN assessed these calculations through ASTEC calculations. Finally, a global agreement was

Data used for safety assessment of reprocessing facilities

International Nuclear Information System (INIS)

Nomura, Yasushi; Suzuki, Atsuyuki; Kanagawa, Akira

1990-08-01

For safety assessment of a reprocessing facility, it is important to know performance of radioactive materials in their accidental release and transfer. Accordingly, it is necessary to collect and prepare data for use in analyses for their performance. In JAERI, experiments such as for data acquisition, for source-term evaluation and for radioactive material transfer, are now planned to be performed. Prior to these experiments, it is decided to investigate data in use for accidental safety assessment of reprocessing plants and their based experimental data, thus to make it possible to recommend reasonable values for safety analysis parameters by evaluating the investigated results, to select the experimental items, to edit a safety assessment handbook and so on. In this line of objectives, JAERI rewarded a two-year contract of investigation to Nuclear Safety Research Association, to make a working group under a special committee on data investigation for reprocessing facility safety assessment. This report is a collection of results reviewed and checked by the working group. The contents consist of two parts, one for investigation and review of data used for safety assessment of domestic or oversea reprocessing facilities, and the other for investigation, review and evaluation of ANSI recommended American standard data reported by E. Walker together with their based experimental data resorting to the original referred reports. (author)

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

International Nuclear Information System (INIS)

Hedin, A.

2008-01-01

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

Procedures for conducting probabilistic safety assessments of nuclear power plants (level 2). Accident progression, containment analysis and estimation of accident source terms

International Nuclear Information System (INIS)

1995-01-01

The present publication on Level 2 PSA is based on a compilation and review of practices in various Member States. It complements Safety Series No. 50-P-4, issued in 1992, on Procedures for Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 1). Refs, figs and tabs

Safety assessment as basis for the decision making process

International Nuclear Information System (INIS)

Ilie, P.; Didita, L.; Danchiv, A.

2005-01-01

This paper deals with the safety assessment for a new near surface repository, particularly for the early stage of repository development using ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) safety assessment methodology. In this stage of the repository life cycle the main purpose of the safety assessment is to demonstrate that the plant is capable to be constructed and operated safely. The paper is based on development of the ASAM (Application of the Safety Assessment Methodologies for Near-Surface Disposal Facilities) Decision Support Subgroup of the Common Aspects Working Group. The implications of decision making for the application of the ISAM methodology on post-closure safety assessment are analysed. Some important elements of the decision-making process with impact on key components of the ISAM process are described. Following the development of Decision Support Subgroup of the ASAM Common Aspects Working Group the proposed change of ISAM methodology is analysed. This approach puts all activities in a decision context where the first iteration of the safety assessment is based on the existing state of knowledge and the initial engineering design. Confidence in the process is accomplished through the direct inclusion of all decision makers and stakeholders in the formulation of decisions, the definition of the state of knowledge, and decision making activities. The decision process is developed in context of undertaking assessments with little site-specific information, this situation is specifically for new planned repository. Limited site-specific information can result in a high degree of uncertainty, therefore it is important first of all to identify the sources of uncertainty arising from the limited nature of the site-specific information and then to apply appropriate approaches to manage the uncertainties and to determine whether the uncertainties are important to the overall safety of the disposal facility

Assessment of safety culture at INPP

International Nuclear Information System (INIS)

Lesin, S.

2002-01-01

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

PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

International Nuclear Information System (INIS)

Paladino, Domenico

2014-01-01

This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

Safety assessment for radioactive waste disposal facility

International Nuclear Information System (INIS)

Thanaletchumy Karuppiah; Mohd Abdul Wahab Yusof; Nik Marzuki Nik Ibrahim; Nurul Wahida Ahmad Khairuddin

2008-08-01

Safety assessments are used to evaluate the performance of a radioactive waste disposal facility and its impact on human health and the environment. This paper presents the overall information and methodology to carry out the safety assessment for a long term performance of a disposal system. A case study was also conducted to gain hands-on experience in the development and justification of scenarios, the formulation and implementation of models and the analysis of results. AMBER code using compartmental modeling approach was used to represent the migration and fate of contaminants in this training. This safety assessment is purely illustrative and it serves as a starting point for each development stage of a disposal facility. This assessment ultimately becomes more detail and specific as the facility evolves. (Author)

Preliminary safety assessment of the WIPP facility

International Nuclear Information System (INIS)

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

1979-01-01

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

LNG Safety Assessment Evaluation Methods

Energy Technology Data Exchange (ETDEWEB)

Muna, Alice Baca [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-05-01

Sandia National Laboratories evaluated published safety assessment methods across a variety of industries including Liquefied Natural Gas (LNG), hydrogen, land and marine transportation, as well as the US Department of Defense (DOD). All the methods were evaluated for their potential applicability for use in the LNG railroad application. After reviewing the documents included in this report, as well as others not included because of repetition, the Department of Energy (DOE) Hydrogen Safety Plan Checklist is most suitable to be adapted to the LNG railroad application. This report was developed to survey industries related to rail transportation for methodologies and tools that can be used by the FRA to review and evaluate safety assessments submitted by the railroad industry as a part of their implementation plans for liquefied or compressed natural gas storage ( on-board or tender) and engine fueling delivery systems. The main sections of this report provide an overview of various methods found during this survey. In most cases, the reference document is quoted directly. The final section provides discussion and a recommendation for the most appropriate methodology that will allow efficient and consistent evaluations to be made. The DOE Hydrogen Safety Plan Checklist was then revised to adapt it as a methodology for the Federal Railroad Administration’s use in evaluating safety plans submitted by the railroad industry.

Promoting and assessment of safety culture within regulatory body

International Nuclear Information System (INIS)

Awasthi, Sumit; Bhattacharya, D.; Koley, J.; Krishnamurthy, P.R.

2015-01-01

Regulators have an important role to play in assisting organizations under their jurisdiction to develop positive safety cultures. It is therefore essential for the regulator to have a robust safety culture as an inherent strategy and communication of this strategy to the organizations it supervises. Atomic Energy Regulatory Board (AERB) emphasizes every utility to institute a good safety culture during various stages of a NPP. The regulatory requirement for establishing organisational safety culture within utility at different stages are delineated in the various AERB safety codes which are presented in the paper. Although the review and assessment of the safety culture is a part of AERB’s continual safety supervision through existing review mechanism, AERB do not use any specific indicators for safety culture assessment. However, establishing and nurturing a good safety culture within AERB helps in encouraging the utility to institute the same. At the induction level AERB provides training to its staffs for regulatory orientation which include a specific course on safety culture. Subsequently, the junior staffs are mentored by seniors while involving them in various regulatory processes and putting them as observers during regulatory decision making process. Further, AERB established a formal procedure for assessing and improving safety culture within its staff as a management system process. The paper describes as a case study the above safety culture assessment process established within AERB

Probabilistic safety assessment for research reactors

International Nuclear Information System (INIS)

1986-12-01

Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

Criticality safety evaluation of Rocky Flats Plant one-gallon shipping containers

Energy Technology Data Exchange (ETDEWEB)

Shaw, M.E.

1991-12-01

Criticality safety calculations have been performed to provide an analytical basis for handling, storage and transport of Rocky Flats Plant (RFP) one-gallon shipping containers. A mass limit was establish for metal (solid uranium or plutonium) and slurries (undissolved U or Pu solids in a mud,'' sludge,'' or slurry''). A separate volume limit was developed for plutonium solutions (liquids, either aqueous or organic, containing no visible undissolved solids).

Safety assessment of radioactive wastes storage 'Mironova Gora'

International Nuclear Information System (INIS)

Serbryakov, B.; Karamushka, V.; Ostroborodov, V.

2000-01-01

A project of transforming the radioactive wastes storage 'Mironova Gora' is under development. A safety assessment of this storage facility was performed to gain assurance on the design decision. The assessment, which was based on the safety assessment methods developed for radioactive wastes repositories, is presented in this paper. (author)

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

International Nuclear Information System (INIS)

Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

2014-01-01

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

2014-10-15

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

Energy Technology Data Exchange (ETDEWEB)

Kumar, Mukesh, E-mail: mukeshd@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarty, Aranyak [School of Nuclear Studies and Application, Jadavpur University, Kolkata 700032 (India); Nayak, A.K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Prasad, Hari; Gopika, V. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

2014-10-15

Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

International Nuclear Information System (INIS)

Kumar, Mukesh; Chakravarty, Aranyak; Nayak, A.K.; Prasad, Hari; Gopika, V.

2014-01-01

Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components

LWR safety studies. Analyses and further assessments relating to the German Risk Assessment Study on Nuclear Power Plants. Vol. 3

International Nuclear Information System (INIS)

1983-01-01

Critical review of the analyses of the German Risk Assessment Study on Nuclear Power Plants (DRS) concerning the reliability of the containment under accident conditions and the conditions of fission product release (transport and distribution in the environment). Main point of interest in this context is an explosion in the steam section and its impact on the containment. Critical comments are given on the models used in the DRS for determining the accident consequences. The analyses made deal with the mathematical models and database for propagation calculations, the methods of dose computation and assessment of health hazards, and the modelling of protective and safety measures. Social impacts of reactor accidents are also considered. (RF) [de

Quality and safety of products containing Ephedra Herba on the Dutch market

OpenAIRE

Lake OA; Slijkhuis C; Maas WF; Vliet MEA van; Kaste D de; Verdonk-Kleinjan W; Keuringsdienst van Waren, regio Zuid; LGO

2001-01-01

We performed analytical studies on dietary supplements and smart products containing Ephedra herba on the Dutch market. Such products are labelled 'from natural, herbal sources' and do not fall under Dutch Medicines Act. Most of the samples tested from 1993 to 1999 contained unacceptably large amounts of ephedrine (EP) alkaloids (the active substances of Ephedra herba) in relation to the safety criteria in the literature. Some samples also contained an effect-enhancing substance (e.g. coffein...

Safety indicators for the safety assessment of radioactive waste disposal. Sixth report of the Working Group on Principles and Criteria for Radioactive Waste Disposal

International Nuclear Information System (INIS)

2003-09-01

The report describes a few indicators that are considered to be the most promising for assessing the long term safety of disposal systems. The safety indicators that are discussed here may be applicable to a range of disposal systems for different waste types, including near surface disposal facilities for low level waste. The appropriateness of the different indicators may, however, vary depending on the characteristics of the waste, the facility and the assessment context. The focus of the report is thus on the use of time-scales of containment and transport, and radionuclide concentrations and fluxes, as indicators of disposal system safety, that may complement the more usual safety indicators of dose and risk. Summarised are the broad elements that a safety case for an underground radioactive waste disposal facility should possess and the role and use of performance and safety indicators within these elements. An overview of performance and safety indicators is given. The use is discussed of dose and risk as safety indicators and, in particular, problems that can arise in their use. Also presented are some specific indicators that have the potential to be used as complementary safety indicators. Discussed is also how fluxes of naturally occurring elements and radionuclides due to the operation of natural processes such as erosion and groundwater discharge may be quantified for comparison with fluxes of waste derived contaminants

Safety assessment of personal care products/cosmetics and their ingredients

International Nuclear Information System (INIS)

Nohynek, Gerhard J.; Antignac, Eric; Re, Thomas; Toutain, Herve

2010-01-01

We attempt to review the safety assessment of personal care products (PCP) and ingredients that are representative and pose complex safety issues. PCP are generally applied to human skin and mainly produce local exposure, although skin penetration or use in the oral cavity, on the face, lips, eyes and mucosa may also produce human systemic exposure. In the EU, US and Japan, the safety of PCP is regulated under cosmetic and/or drug regulations. Oxidative hair dyes contain arylamines, the most chemically reactive ingredients of PCP. Although arylamines have an allergic potential, taking into account the high number of consumers exposed, the incidence and prevalence of hair dye allergy appears to be low and stable. A recent (2001) epidemiology study suggested an association of oxidative hair dye use and increased bladder cancer risk in consumers, although this was not confirmed by subsequent or previous epidemiologic investigations. The results of genetic toxicity, carcinogenicity and reproductive toxicity studies suggest that modern hair dyes and their ingredients pose no genotoxic, carcinogenic or reproductive risk. Recent reports suggest that arylamines contained in oxidative hair dyes are N-acetylated in human or mammalian skin resulting in systemic exposure to traces of detoxified, i.e. non-genotoxic, metabolites, whereas human hepatocytes were unable to transform hair dye arylamines to potentially carcinogenic metabolites. An expert panel of the International Agency on Research of Cancer (IARC) concluded that there is no evidence for a causal association of hair dye exposure with an elevated cancer risk in consumers. Ultraviolet filters have important benefits by protecting the consumer against adverse effects of UV radiation; these substances undergo a stringent safety evaluation under current international regulations prior to their marketing. Concerns were also raised about the safety of solid nanoparticles in PCP, mainly TiO 2 and ZnO in sunscreens. However

Safety assessment of personal care products/cosmetics and their ingredients.

Science.gov (United States)

Nohynek, Gerhard J; Antignac, Eric; Re, Thomas; Toutain, Herve

2010-03-01

We attempt to review the safety assessment of personal care products (PCP) and ingredients that are representative and pose complex safety issues. PCP are generally applied to human skin and mainly produce local exposure, although skin penetration or use in the oral cavity, on the face, lips, eyes and mucosa may also produce human systemic exposure. In the EU, US and Japan, the safety of PCP is regulated under cosmetic and/or drug regulations. Oxidative hair dyes contain arylamines, the most chemically reactive ingredients of PCP. Although arylamines have an allergic potential, taking into account the high number of consumers exposed, the incidence and prevalence of hair dye allergy appears to be low and stable. A recent (2001) epidemiology study suggested an association of oxidative hair dye use and increased bladder cancer risk in consumers, although this was not confirmed by subsequent or previous epidemiologic investigations. The results of genetic toxicity, carcinogenicity and reproductive toxicity studies suggest that modern hair dyes and their ingredients pose no genotoxic, carcinogenic or reproductive risk. Recent reports suggest that arylamines contained in oxidative hair dyes are N-acetylated in human or mammalian skin resulting in systemic exposure to traces of detoxified, i.e. non-genotoxic, metabolites, whereas human hepatocytes were unable to transform hair dye arylamines to potentially carcinogenic metabolites. An expert panel of the International Agency on Research of Cancer (IARC) concluded that there is no evidence for a causal association of hair dye exposure with an elevated cancer risk in consumers. Ultraviolet filters have important benefits by protecting the consumer against adverse effects of UV radiation; these substances undergo a stringent safety evaluation under current international regulations prior to their marketing. Concerns were also raised about the safety of solid nanoparticles in PCP, mainly TiO(2) and ZnO in sunscreens. However

Uncertainty estimation in nuclear power plant probabilistic safety assessment

International Nuclear Information System (INIS)

Guarro, S.B.; Cummings, G.E.

1989-01-01

Probabilistic Risk Assessment (PRA) was introduced in the nuclear industry and the nuclear regulatory process in 1975 with the publication of the Reactor Safety Study by the U.S. Nuclear Regulatory Commission. Almost fifteen years later, the state-of-the-art in this field has been expanded and sharpened in many areas, and about thirty-five plant-specific PRAs (Probabilistic Risk Assessments) have been performed by the nuclear utility companies or by the U.S. Nuclear Regulatory commission. Among the areas where the most evident progress has been made in PRA and PSA (Probabilistic Safety Assessment, as these studies are more commonly referred to in the international community outside the U.S.) is the development of a consistent framework for the identification of sources of uncertainty and the estimation of their magnitude as it impacts various risk measures. Techniques to propagate uncertainty in reliability data through the risk models and display its effect on the top level risk estimates were developed in the early PRAs. The Seismic Safety Margin Research Program (SSMRP) study was the first major risk study to develop an approach to deal explicitly with uncertainty in risk estimates introduced not only by uncertainty in component reliability data, but by the incomplete state of knowledge of the assessor(s) with regard to basic phenomena that may trigger and drive a severe accident. More recently NUREG-1150, another major study of reactor risk sponsored by the NRC, has expanded risk uncertainty estimation and analysis into the realm of model uncertainty related to the relatively poorly known post-core-melt phenomena which determine the behavior of the molten core and of the rector containment structures

Assessment of impact of construction materials on the ecological safety of home

Directory of Open Access Journals (Sweden)

Zhigulina Anna

2017-01-01

Full Text Available The article deals with the problems of creating environmentally friendly aerial environment within residential premises. The main sources causing air pollution of urban housing are determined and classified. The origins of air pollution sources of residential premises are adopted as the classifying criterion. The sources of contamination are defined and assessed. Particular attention is paid to the choice of environmentally friendly building materials. The methodology for assessing toxicity of industrial waste used in the production of housing materials is developed to assess the comfort and environmental safety of home. The idea of creating “Residential buldings ID” containing information on the construction materials used is introduced.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-30

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

Self-assessment of operational safety for nuclear power plants

International Nuclear Information System (INIS)

1999-12-01

Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and the safety culture as a whole. Although the primary beneficiaries of the self-assessment process are the plant and operating organization, the results of the self-assessments are also used, for example, to increase the confidence of the regulator in the safe operation of an installation, and could be used to assist in meeting obligations under the Convention on Nuclear Safety. Such considerations influence the form of assessment, as well as the type and detail of the results. The concepts developed in this report present the basic approach to self-assessment, taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject. This report will be used in IAEA sponsored workshops and seminars on operational safety that include the topic of self-assessment

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Analysis of truncation limit in probabilistic safety assessment

International Nuclear Information System (INIS)

Cepin, Marko

2005-01-01

A truncation limit defines the boundaries of what is considered in the probabilistic safety assessment and what is neglected. The truncation limit that is the focus here is the truncation limit on the size of the minimal cut set contribution at which to cut off. A new method was developed, which defines truncation limit in probabilistic safety assessment. The method specifies truncation limits with more stringency than presenting existing documents dealing with truncation criteria in probabilistic safety assessment do. The results of this paper indicate that the truncation limits for more complex probabilistic safety assessments, which consist of larger number of basic events, should be more severe than presently recommended in existing documents if more accuracy is desired. The truncation limits defined by the new method reduce the relative errors of importance measures and produce more accurate results for probabilistic safety assessment applications. The reduced relative errors of importance measures can prevent situations, where the acceptability of change of equipment under investigation according to RG 1.174 would be shifted from region, where changes can be accepted, to region, where changes cannot be accepted, if the results would be calculated with smaller truncation limit

Risk assessment of safety violations for coal mines

Energy Technology Data Exchange (ETDEWEB)

Megan Orsulaka; Vladislav Kecojevicb; Larry Graysona; Antonio Nietoa [Pennsylvania State University, University Park, PA (United States). Dept of Energy and Mineral Engineering

2010-09-15

This article presents an application of a risk assessment approach in characterising the risks associated with safety violations in underground bituminous mines in Pennsylvania using the Mine Safety and Health Administration (MSHA) citation database. The MSHA database on citations provides an opportunity to assess risks in mines through scrutiny of violations of mandatory safety standards. In this study, quantitative risk assessment is performed, which allows determination of the frequency of occurrence of safety violations (through associated citations) as well as the consequences of them in terms of penalty assessments. Focus is on establishing risk matrices on citation experiences of mines, which can give early indication of emerging potentially serious problems. The resulting frequency, consequence and risk rankings present valuable tools for prioritising resource allocations, determining control strategies, and could potentially contribute to more proactive prevention of incidents and injuries.

Healthcare professionals’ views of feedback on patient safety culture assessment.

OpenAIRE

Zwijnenberg, N.C.; Hendriks, M.; Hoogervorst-Schilp, J.; Wagner, C.

2016-01-01

Background: By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals’ views on the feedback of a patient safety culture assessment. Methods: Twenty four hospitals participated in a patient safety culture assessment in 2012. Hospital departments received feedback in a report and on a web...

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

International Nuclear Information System (INIS)

1994-01-01

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

Methods for assessing NPP containment pressure boundary integrity

International Nuclear Information System (INIS)

Naus, D.J.; Ellingwood, B.R.; Graves, H.L.

2004-01-01

Research is being conducted to address aging of the containment pressure boundary in light-water reactor plants. Objectives of this research are to (1) understand the significant factors relating to corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and of liners of concrete containments; (2) provide the U.S. Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation. Activities include development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of candidate techniques for inspection of inaccessible regions of containment metallic pressure boundaries; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion

Safety Assessment of Polyether Lanolins as Used in Cosmetics.

Science.gov (United States)

Becker, Lillian C; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan; Heldreth, Bart

The Cosmetic Ingredient Review (CIR) Expert Panel (Panel) assessed the safety of 39 polyether lanolin ingredients as used in cosmetics. These ingredients function mostly as hair conditioning agents, skin conditioning agent-emollients, and surfactant-emulsifying agents. The Panel reviewed available animal and clinical data, from previous CIR safety assessments of related ingredients and components. The similar structure, properties, functions, and uses of these ingredients enabled grouping them and using the available toxicological data to assess the safety of the entire group. The Panel concluded that these polyether lanolin ingredients are safe in the practices of use and concentration as given in this safety assessment.

Hydrogen related safety issues in the context of containments of Indian PHWRs

International Nuclear Information System (INIS)

Markendeya, S.G.; Ghosh, A.K.; Kushwaha, H.S.; Venkat Raj, V.

2002-01-01

Full text: Assessment of risk due to hydrogen released during postulated hypothetical accident scenarios in the nuclear power plants (NPPs) has been an important area of R and D studies world over. The issues, such as appropriate methodologies for estimation of hydrogen source term and for hydrogen dispersion calculations, technology development for hydrogen mitigation in containment of NPPs and assessment of damage due to deflagration/detonation of hydrogen (if it occurs) are being addressed as a part of some of the multidisciplinary study programs currently underway in BARC. While a significant overall progress has been achieved in general as a result of these programs, requirements of further fine-tuning of these studies have also emerged. The present paper takes a brief look at the current state-of the-art technology available to address these issues. The progress of R and D studies underway at BARC has also been critically reviewed in the paper to bring out necessary planning of further studies so as to enhance the safety of Indian NPPs

Probabilistic safety assessment as a standpoint for decision making

International Nuclear Information System (INIS)

Cepin, M.

2001-01-01

This paper focuses on the role of probabilistic safety assessment in decision-making. The prerequisites for use of the results of probabilistic safety assessment and the criteria for the decision-making based on probabilistic safety assessment are discussed. The decision-making process is described. It provides a risk evaluation of impact of the issue under investigation. Selected examples are discussed, which highlight the described process. (authors)

Safety Assessment of Multi Purpose Small Payload Rack(MSPR)

Science.gov (United States)

Mizutani, Yoshinobu; Takada, Satomi; Murata, Kosei; Ozawa, Daisaku; Kobayashi, Ryoji; Nakamura, Yasuhiro

2010-09-01

We are reporting summary of preliminary safety assessment for Multi Purpose Small Payload Rack(MSPR), which is one of the micro gravity experiment facilities that are being developed for the 2nd phase JEM utilization(JEM: Japanese Experiment Module) that will be launched on H-II Transfer Vehicle(HTV) 2nd flight in 2011. MSPR is used for multi-purpose micro-g experiment providing experimental spaces and work stations. MSPR has three experimental spaces; first, there is a space called Work Volume(WV) with capacity volume of approximately 350 litters, in which multiple resources including electricity, communication, and moving image functions can be used. Within this space, installation of devices can be done by simple, prompt attachment by Velcro and pins with high degree of flexibility. Second, there is Small Experiment Area(SEA), with capacity volume of approximately 70 litters, in which electricity, communication, and moving image functions can also be used in the same way as WV. These spaces protect experiment devices and specimens from contingent loads by the crewmembers. Third, there is Work Bench with area of 0.5 square meters, on which can be used for maintenance, inspection and data operations of installed devices, etc. This bench can be stored in the rack during contingency. Chamber for Combustion Experiment(CCE) that is planned to be installed in WV is a pressure-resistant experimental container that can be used to seal hazardous materials from combustion experiments. This CCE has double sealing design in chamber itself, which resist gas leakage under normal the temperature and pressure. Electricity, communication, moving image function can be used in the same way as WV. JAXA Phase 2 Safety Review Panel(SRP) has been held in April, 2010. For safety analysis of MSPR, hazards were identified based on Fault Tree Analysis methodology and then these hazards were classified into either eight ISS standard-type hazards or eight unique-type hazards that requires

The Safety Assessment Framework Tool (SAFRAN) - Description, Overview and Applicability

International Nuclear Information System (INIS)

Alujevic, Luka

2014-01-01

The SAFRAN tool (Safety Assessment Framework) is a user-friendly software application that incorporates the methodologies developed in the SADRWMS (Safety Assessment Driven Radioactive Waste Management Solutions) project. The International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of all types of radioactive waste, including disused sources, small volumes, legacy and decommissioning waste, operational waste, and large volume naturally occurring radioactive material residues. SAFRAN provides aid in: Describing the predisposal RW management activities in a systematic way, Conducting the SA (safety assessment) with clear documentation of the methodology, assumptions, input data and models, Establishing a traceable and transparent record of the safety basis for decisions on the proposed RW management solutions, Demonstrating clear consideration of and compliance with national and international safety standards and recommendations. The SAFRAN tool allows the user to visibly, systematically and logically address predisposal radioactive waste management and decommissioning challenges in a structured way. It also records the decisions taken in such a way that it constitutes a justifiable safety assessment of the proposed management solutions. The objective of this paper is to describe the SAFRAN architecture and features, properly define the terms safety case and safety assessment, and to predict the future development of the SAFRAN tool and assess its applicability to the construction of a future LILW (Low and Intermediate Level Waste) storage facility and repository in Croatia, taking into account all the capabilities and modelling features of the SAFRAN tool. (author)

Safety factors for neutron fluences in NPP safety assessment

International Nuclear Information System (INIS)

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

2016-01-01

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

Ensuring the quality of occupational safety risk assessment.

Science.gov (United States)

Pinto, Abel; Ribeiro, Rita A; Nunes, Isabel L

2013-03-01

In work environments, the main aim of occupational safety risk assessment (OSRA) is to improve the safety level of an installation or site by either preventing accidents and injuries or minimizing their consequences. To this end, it is of paramount importance to identify all sources of hazards and assess their potential to cause problems in the respective context. If the OSRA process is inadequate and/or not applied effectively, it results in an ineffective safety prevention program and inefficient use of resources. An appropriate OSRA is an essential component of the occupational safety risk management process in industries. In this article, we performed a survey to elicit the relative importance for identified OSRA tasks to enable an in-depth evaluation of the quality of risk assessments related to occupational safety aspects on industrial sites. The survey involved defining a questionnaire with the most important elements (tasks) for OSRA quality assessment, which was then presented to safety experts in the mining, electrical power production, transportation, and petrochemical industries. With this work, we expect to contribute to the main question of OSRA in industries: "What constitutes a good occupational safety risk assessment?" The results obtained from the questionnaire showed that experts agree with the proposed OSRA process decomposition in steps and tasks (taxonomy) and also with the importance of assigning weights to obtain knowledge about OSRA task relevance. The knowledge gained will enable us, in the near future, to build a framework to evaluate OSRA quality for industrial sites. © 2012 Society for Risk Analysis.

Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

International Nuclear Information System (INIS)

Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

2008-01-01

The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design

Exploiting data from safety investigations and processes to assess performance of safety management aspects

NARCIS (Netherlands)

Karanikas, Nektarios

2016-01-01

This paper presents an alternative way to use records from safety investigations as a means to support the evaluation of safety management (SM) aspects. Datasets from safety investigation reports and progress records of an aviation organization were analyzed with the scope of assessing safety

NPP Krsko periodic safety review. Safety assessment and analyses

International Nuclear Information System (INIS)

Basic, I.; Spiler, J.; Thaulez, F.

2002-01-01

Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

Safety assessment of a lithium target

International Nuclear Information System (INIS)

Burgazzi, Luciano; Roberta, Ferri; Barbara, Giannone

2006-01-01

This paper addresses the safety assessment of the lithium target of the International Fusion Materials Irradiation Facility (IFMIF) through evaluating the most important risk factors related to system operation and verifying the fulfillment of the safety criteria. The hazard assessment is based on using a well-structured Failure Mode and Effect Analysis (FMEA) procedure by detailing on a component-by-component basis all the possible failure modes and identifying their effects on the plant. Additionally, a systems analysis, applying the fault tree technique, is performed in order to evaluate, from a probabilistic standpoint, all the relevant and possible failures of each component required for safe system operation and assessing the unavailability of the lithium target system. The last task includes the thermal-hydraulic transient analysis of the target lithium loop, including operational and accident transients. A lithium target loop model is developed, using the RELAP5/Mod3.2 thermal-hydraulic code, which has been modified to include specific features of IFMIF itself. The main conclusions are that target safety is fulfilled, the hazards associated with lithium operation are confined within the IFMIF security boundaries, the environmental impact is negligible, and the plant responds to the simulated transients by being able to reach steady conditions in a safety situation

The role of natural analogues in safety assessment and acceptability

International Nuclear Information System (INIS)

Papp, Toenis

1987-01-01

The safety assessment must evaluate the level of safety for a repository, the confidence that can be placed on the assessment and how well the repository can meet the acceptance criteria of the society. Many of the processes and phenomena that govern the long term performance of a deep geologic repository for radioactive waste also take place in nature. To investigate these natural analogues and try to validate the models on which the safety assessment are based is a main task in the effort to build of confidence in the safety assessments. The assessment of the safety of a repository can, however, not only be based on good models. The possible role of natural analogues or natural evidence in other parts of the safety assessment is discussed. Specially with regard to - the need to demonstrate that all relevant processes have been taken into account, and that the important ones have been validated to an acceptable level for relevant parameters spans, -the definition and analysis of external scenarios for the safety assessment and for the claim that all reasonable scenarios have been addressed, - the public confidence in the long-term relevance of the acceptance criteria. (author)

MAPLE research reactor safety uncertainty assessment methodology

International Nuclear Information System (INIS)

Sills, H.E.; Duffey, R.B.; Andres, T.H.

1999-01-01

The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

International Nuclear Information System (INIS)

Armstrong, D.L.

1994-08-01

This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment

Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Armstrong, D.L.

1994-08-01

This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment.

Assessment of gas flammability in transuranic waste container

International Nuclear Information System (INIS)

Connolly, M.J.; Loehr, C.A.; Djordjevic, S.M.; Spangler, L.R.

1995-01-01

The Safety Analysis Report for the TRUPACT-II Shipping Package [Transuranic Package Transporter-II (TRUPACT-II) SARP] set limits for gas generation rates, wattage limits, and flammable volatile organic compound (VOC) concentrations in transuranic (TRU) waste containers that would be shipped to the Waste Isolation Pilot Plant (WIPP). Based on existing headspace gas data for drums stored at the Idaho National Engineering Laboratory (INEL) and the Rocky Flats Environmental Technology Site (RFETS), over 30 percent of the contact-handled TRU waste drums contain flammable VOC concentrations greater than the limit. Additional requirements may be imposed for emplacement of waste in the WIPP facility. The conditional no-migration determination (NMD) for the test phase of the facility required that flame tests be performed if significant levels of flammable VOCs were present in TRU waste containers. This paper describes an approach for investigating the potential flammability of TRU waste drums, which would increase the allowable concentrations of flammable VOCS. A flammability assessment methodology is presented that will allow more drums to be shipped to WIPP without treatment or repackaging and reduce the need for flame testing on drums. The approach includes experimental work to determine mixture lower explosive limits (MLEL) for the types of gas mixtures observed in TRU waste, a model for predicting the MLEL for mixtures of VOCS, hydrogen, and methane, and revised screening limits for total flammable VOCs concentrations and concentrations of hydrogen and methane using existing drum headspace gas data and the model predictions

Nuclear criticality safety calculational analysis for small-diameter containers

International Nuclear Information System (INIS)

LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.

1995-11-01

This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can (open-quotes F-canclose quotes); (4) 5.25-inch-diameter by 15-inch-high steel can (open-quotes Z-canclose quotes); and (5) 5.0-inch-diameter by 9-inch-high polybottle (open-quotes CO-4close quotes). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO 2 F 2 +H 2 O and UF 4 +oil materials at 100% and 10% enrichments and U 3 O 8 , and H 2 O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k eff + 2σ < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant

RSAS: a Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Dixon, B.W.; Bray, M.A.

1985-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

An assessment of the impact of home safety assessments on fires and fire-related injuries: a case study of Cheshire Fire and Rescue Service.

Science.gov (United States)

Arch, B N; Thurston, M N

2013-06-01

Deaths and injuries related to fires are largely preventable events. In the UK, a plethora of community-based fire safety initiatives have been introduced over the last 25 years, often led by fire and rescue services, to address this issue. This paper focuses on one such initiative--home safety assessments (HSAs). Cheshire Fire and Rescue Service (in England) implemented a uniquely large-scale HSA intervention. This paper assesses its effectiveness. The impact of HSAs was assessed in relation to three outcomes: accidental dwelling fires (ADFs), ADFs contained and injuries arising from ADFs. A two-period comparison in fire-related rates of incidences in Cheshire between 2002 and 2011 was implemented, using Poisson regression and adjusting for the national temporal trend using a control group comprising the 37 other English non-metropolitan fire-services. Significant reductions were observed in rates of ADFs [incidence rate ratios (IRR): 0.79, 95% confidence interval (CI): 0.74-0.83, P fires contained to room of origin. There is strong evidence to suggest that the intervention was successful in reducing domestic fires and related injuries.

Safety assessment and detection methods of genetically modified organisms.

Science.gov (United States)

Xu, Rong; Zheng, Zhe; Jiao, Guanglian

2014-01-01

Genetically modified organisms (GMOs), are gaining importance in agriculture as well as the production of food and feed. Along with the development of GMOs, health and food safety concerns have been raised. These concerns for these new GMOs make it necessary to set up strict system on food safety assessment of GMOs. The food safety assessment of GMOs, current development status of safety and precise transgenic technologies and GMOs detection have been discussed in this review. The recent patents about GMOs and their detection methods are also reviewed. This review can provide elementary introduction on how to assess and detect GMOs.

Defining initiating events for purposes of probabilistic safety assessment

International Nuclear Information System (INIS)

1993-09-01

This document is primarily directed towards technical staff involved in the performance or review of plant specific Probabilistic Safety Assessment (PSA). It highlights different approaches and provides typical examples useful for defining the Initiating Events (IE). The document also includes the generic initiating event database, containing about 300 records taken from about 30 plant specific PSAs. In addition to its usefulness during the actual performance of a PSA, the generic IE database is of the utmost importance for peer reviews of PSAs, such as the IAEA's International Peer Review Service (IPERS) where reference to studies on similar NPPs is needed. 60 refs, figs and tabs

Guidelines for Self-assessment of Research Reactor Safety

International Nuclear Information System (INIS)

2018-01-01

Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

Radiation risk assessment for the transport of radioisotopes using KRI-BGM B(U) type container

International Nuclear Information System (INIS)

Cho, Woon-Kap

2008-01-01

The radiation risks were estimated for the transportation of radioisotopes using KRI-BGM transport container. KRI-BGM container was specially designed for transportation of large amount of radioisotopes for industrial or medical applications. The container can carry maximum 370 TBq of solid Ir-192, 29.6 TBq of liquid Mo-99 and 37 TBq of liquid I-131 respectively. For the radiation risk assessment, it was assumed that maximum design activity of those radioisotopes was transported. Transportation route is from Daejeon where radioisotopes are produced to Seoul where radioisotopes are consumed. Transport distance is 200 km including highway and downtown area from Daejeon to Seoul. As the transportation conveyance, an ordinary cargo truck is used exclusively. Radiation risks were estimated for incident free and accident condition of transportation and RADTRAN 5.6 was used as the risk assessment tool. For the risk assessment of radioisotopes transportation, various parameters such as population density around transport route, weather condition, probability of specific accidents such as impact, fire, etc. were considered. From the results of this study, the exclusive transportation of radioisotopes using KRI-BGM transport container by truck showed low radiological risks with manageable safety and health consequences. This paper discusses the methods and results of the radiation risks assessment for the radioisotopes transportation by an ordinary truck and presents the expected radiation risks in person-Sv and latent cancer fatalities. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

ASCOT guidelines revised 1996 edition. Guidelines for organizational self-assessment of safety culture and for reviews by the assessment of safety culture in organizations team

International Nuclear Information System (INIS)

1996-01-01

In order to properly assess safety culture, it is necessary to consider the contribution of all organizations which have an impact on it. Therefore, while assessing the safety culture in an operating organization it is necessary to address at least its interfaces with the local regulatory agency, utility corporate headquarters and supporting organizations. These guidelines are primarily intended for use by any organization wishing to conduct a self-assessment of safety culture. They should also serve as a basis for conducting an international peer review of the organization's self-assessment carried out by an ASCOT (Assessment of Safety Culture in Organizations Team) mission

An Assessment of the Fire Safety Hazard Associated with External Fire Spread in Tall Buildings with Combustible Façade Material

DEFF Research Database (Denmark)

Lavard Brogaard, Nicholas; Torero, Jose L.; Jomaas, Grunde

2014-01-01

in order to obtain a conclusive assessment of the fire safety hazards associated with combustible facades. Prescriptive fire safety codes are typically not allowing any type of combustible façade in buildings that are taller than 2-3 stories. However, a performance based approach does not contain height...

Safety assessment of OPG's used fuel for dry storage

International Nuclear Information System (INIS)

Roman, H.; Khan, A.

2005-01-01

'Full text:' Ontario Power Generation (OPG) operates the Pickering Waste Management Facility (PWMF) and Western Waste Management Facility (WWMF) where OPG has been storing 10-year or older used fuel in the Dry Storage Containers (DSCs) since 1996 and 2003 respectively. The construction licence for the Darlington Used Fuel Dry Storage Facility (DUFDSF) was obtained in August 2004. Safety assessment of the used fuel for dry storage is required to support each request for regulatory approval to construct and operate a dry storage facility. The objective of the safety assessment is to assess the used fuel performance under normal operation and postulated credible accident scenarios. A reference used fuel bundle is defined based on the operating history and data on fuel discharged from the reactors of the specific nuclear generating station. The characteristics of the reference used fuel bundle are used to calculate the nuclide inventory, source term and decay heat used for the assessment. When assessing malfunctions and accidents, postulated external and internal events are considered. Consideration is also given to the design basis accidents of the specific nuclear generating station that could affect the used fuel under dry storage. For those events deemed credible (i.e. probability > 10 -7 ), a bounding fuel failure consequence is predicted. Given the chemical characteristics of the radionuclides in used fuel, the design of the CANDU fuel and the conditions inside the DSC, in the event that a used fuel bundle should become damaged during used fuel dry storage operations, the only significant radionuclides species that are volatile are krypton-85 and tritium. Release of these radionuclides is considered in calculating public and worker doses. (author)

Application of fuzzy set theory for safety culture and safety management assessment of Kartini research reactor

International Nuclear Information System (INIS)

Syarip; Hauptmanns, U.

2000-01-01

The safety culture status of nuclear power plant is usually assessed through interview and/or discussions with personnel and management in plant, and an assessment of the pertinent documentation. The approach for safety culture assessment described in IAEA Safety Series, make uses of a questionnaire composed of questions which require 'Yes' or 'No' as an answer. Hence, it is basically a check-list approach which is quite common for safety assessments in industry. Such a procedure ignores the fact that the expert answering the question usually has knowledge which goes far beyond a mere binary answer. Additionally, many situations cannot readily be described in such restricted terms. Therefore, it was developed a checklist consisting of questions which are formulated such that they require more than a simple 'yes' or 'no' as an answer. This allows one to exploit the expert knowledge of the analyst appropriately by asking him to qualify the degree of compliance of each of the topics examined. The method presented has proved useful in assessing the safety culture and quality of safety management of the research reactor. The safety culture status and the quality of safety management of Kartini research reactor is rated as 'average'. The method is also flexible and allows one to add questions to existing areas or to introduce new areas covering related topics

Choice and complexation of techniques and tools for assessment of NPP I and C systems safety

International Nuclear Information System (INIS)

Illiashenko, Oleg; Babeshko, Eugene

2011-01-01

There are a lot of techniques to analyze and assess reliability and safety of NPP Instrumentation and Control (I and C) systems (e.g. FMEA - Failure Modes and Effects Analysis and its modifications, FTA - Fault Tree Analysis, HAZOP - Hazard and Operability Analysis, RBD - Reliability Block Diagram, Markov Models, etc.) and quantity of tools based on these techniques is constantly increasing. Known ways of safety assessment, as well as problems of their choice and complexation are analyzed. Objective of the paper is the development of general 'technique of techniques choosing' and tool for support of such technique. The following criteria are used for analysis and comparison and their features are described: compliance to normative documents; experience of application in industry; methods used for assessment of system NPP I and C safety; tool architecture/framework; reporting; vendor support, etc. Comparative analysis results of existing T and T - Tools and Techniques for safety analysis are presented in matrix form ('Tools-Criterion') with example. Features of complexation of different safety assessment techniques (FMECA, FTA, RBD, Markov Models) are described. The proposed technique is implemented as special tool for decision-making. The proposed technique was used for development of RPC Radiy company standard CS 66. This guide contains requirements and procedures of FMECA analysis of developed and produced NPP I and C systems based on RADIY platform. (author)

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

International Nuclear Information System (INIS)

Lee, Seunghee; Kim, Juyoul

2017-01-01

Highlights: • Post-closure safety assessment of near surface disposal facility for DSRS was performed. • Engineered vault and rock-cavern type were considered for normal and well scenario. • 14 C, 226 Ra, 241 Am were primary nuclides contributing large portion of exposure dose. • Near surface disposal of DSRSs containing 14 C, 226 Ra and 241 Am should be restricted. - Abstract: Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1 × 10 −7 mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1 mSv/yr. However, in the

Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

International Nuclear Information System (INIS)

Williams, R.A.

1995-01-01

The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

IRSN-ANCCLI partnership. Work session on Complementary safety assessments - November 2011

International Nuclear Information System (INIS)

Lachaume, Jean-Luc; Lheureux, Yves; Sene, Monique; Sene, Raymond; Jorel, Martial; Lavarenne, Caroline; Rousseau, Jean-Marie; Rebour, Vincent; Baumont, David; Dupuy, Patricia

2011-11-01

After an overview by the ASN of complementary safety assessments and an assessment of 'post-Fukushima' inspections of basic nuclear installations, the contributions (Power Point presentations) of this seminar proposed: the opinion of the Gravelines CLI (local information commission) on the Gravelines complementary safety assessment report, an analysis and discussion by the GSIEN on reports of complementary assessment of safety of nuclear installations with respect to the Fukushima accident, an analysis by the IRSN of complementary safety assessments performed by operators, the IRSN approach to analyze complementary safety assessments, reports on installation conditions, external flooding and seismic hazard, 'meltdown prevention' aspects in the management of accidental situations in EDF reactors

Assessment of the effect of nitrogen gas on passive containment cooling system performance

International Nuclear Information System (INIS)

Ha, Huiun; Suh, Jungsoo

2016-01-01

As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR

Assessment of the effect of nitrogen gas on passive containment cooling system performance

Energy Technology Data Exchange (ETDEWEB)

Ha, Huiun; Suh, Jungsoo [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR.

A review on safety and efficacy of products containing Longifolia

Directory of Open Access Journals (Sweden)

Abdul Hafeez Ahmad Hamdi

2016-01-01

Full Text Available Eurycoma longifolia (commonly called tongkat ali is a flowering plant in the family Simaroubaceae, native to Indonesia, Malaysia, and, to a lesser extent, Thailand, Vietnam and also Laos. The roots extract of E. longifolia, is a well-known traditional herbal medicine in Asia used for many purposes such as sexual dysfunction, aging, malaria, cancer, diabetes, anxiety, aches, constipation, exercise recovery, fever, increased energy, increased strength, leukemia, osteoporosis, stress, syphilis and glandular swelling. The roots are also used as an aphrodisiac, antibiotic, appetite stimulant and health supplement. It is very important to conserve this valuable medicinal plant for the health benefit of future generations. The purpose of this review article is to evaluate and summarize the existing literatures regarding the efficacy and safety of products which contain E. longifolia as its main ingredient. In summary, based on the literature evaluated in this review article, products which contain tongkat ali showed a clinical benefit on improving erectile dysfunction as well as a good safety profile. We recommend consumers to check the level of the bioactive compound “eurycomanone” as their guide before choosing any E. longifolia product.

Initial development of a practical safety audit tool to assess fleet safety management practices.

Science.gov (United States)

Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

2012-07-01

Work-related vehicle crashes are a common cause of occupational injury. Yet, there are few studies that investigate management practices used for light vehicle fleets (i.e. vehicles less than 4.5 tonnes). One of the impediments to obtaining and sharing information on effective fleet safety management is the lack of an evidence-based, standardised measurement tool. This article describes the initial development of an audit tool to assess fleet safety management practices in light vehicle fleets. The audit tool was developed by triangulating information from a review of the literature on fleet safety management practices and from semi-structured interviews with 15 fleet managers and 21 fleet drivers. A preliminary useability assessment was conducted with 5 organisations. The audit tool assesses the management of fleet safety against five core categories: (1) management, systems and processes; (2) monitoring and assessment; (3) employee recruitment, training and education; (4) vehicle technology, selection and maintenance; and (5) vehicle journeys. Each of these core categories has between 1 and 3 sub-categories. Organisations are rated at one of 4 levels on each sub-category. The fleet safety management audit tool is designed to identify the extent to which fleet safety is managed in an organisation against best practice. It is intended that the audit tool be used to conduct audits within an organisation to provide an indicator of progress in managing fleet safety and to consistently benchmark performance against other organisations. Application of the tool by fleet safety researchers is now needed to inform its further development and refinement and to permit psychometric evaluation. Copyright © 2012 Elsevier Ltd. All rights reserved.

Animal-Free Chemical Safety Assessment

Directory of Open Access Journals (Sweden)

George D Loizou

2016-07-01

Full Text Available The exponential growth of the Internet of Things and the global popularity and remarkable decline in cost of the mobile phone is driving the digital transformation of medical practice. The rapidly maturing digital, nonmedical world of mobile (wireless devices, cloud computing and social networking is coalescing with the emerging digital medical world of omics data, biosensors and advanced imaging which offers the increasingly realistic prospect of personalized medicine. Described as a potential seismic shift from the current healthcare model to a wellness paradigm that is predictive, preventative, personalized and participatory, this change is based on the development of increasingly sophisticated biosensors which can track and measure key biochemical variables in people. Additional key drivers in this shift are metabolomic and proteomic signatures, which are increasingly being reported as pre-symptomatic, diagnostic and prognostic of toxicity and disease. These advancements also have profound implications for toxicological evaluation and safety assessment of pharmaceuticals and environmental chemicals. An approach based primarily on human in vivo and high-throughput in vitro human cell-line data is a distinct possibility. This would transform current chemical safety assessment practise which operates in a human data poor to a human data rich environment. This could also lead to a seismic shift from the current animal-based to an animal-free chemical safety assessment paradigm.

Additional safety assessments. Report by the Nuclear Safety Authority - December 2011

International Nuclear Information System (INIS)

2011-12-01

The first part of this voluminous report proposes an assessment of targeted audits performed in French nuclear installations (water pressurized reactors on the one hand, laboratories, factories and waste and dismantling installations on the other hand) on issues related to the Fukushima accident. The examined issues were the protection against flooding and against earthquake, and the loss of electricity supplies and of cooling sources. The second part addresses the additional safety assessments of the reactors and the European resistance tests: presentation of the French electronuclear stock, earthquake, flooding and natural hazards (installation sizing, safety margin assessment), loss of electricity supplies and cooling systems, management of severe accidents, subcontracting conditions. The third part addresses the same issues for nuclear installations other than nuclear power reactors

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

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

Device for removing hydrogen gas from the safety containment vessel of a nuclear reactor

International Nuclear Information System (INIS)

Stiefel, M.

1983-01-01

The safe processing of all concentrations of gas mixtures should be possible with such a device using a thermal recombiner of compact construction. A recombiner consisting of a metal case and diverter sheets situated in it is heated by induction. The incoming pipe for the gas mixture enriched with hydrogen and the outgoing pipe for the gas mixture with low hydrogen content are connected together by a three way valve. The third connection to the safety valve takes the larger port of the gas mixture with low hydrogen content back to the safety containment vessel. Sufficient amount of the gas mixture with low hydrogen content is taken via the three way valve to the safety containment vessel to ensure that the hydrogen content of the gas mixture taken to the recombiner remains below the 4% by volume limit. (orig./PW)

Radiation safety of sealed sources and equipment containing them

International Nuclear Information System (INIS)

1993-01-01

The guide gives information and requirements concerning the technical construction, installation, use and licensing of devices containing sealed radioactive sources in order to ensure the operational safety. The requirements are in accordance with the international standards ISO 1677, ISO 2919, ISO 7205 and Nordic Recommendations on radiation protection for radionuclide gauges in permanent installation. The guide explains also the practical measures that must be taken into account when a radiation device is repaired, maintained or removed from the use. (8 refs.)

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

Reliability-based condition assessment of steel containment and liners

International Nuclear Information System (INIS)

Ellingwood, B.; Bhattacharya, B.; Zheng, R.

1996-11-01

Steel containments and liners in nuclear power plants may be exposed to aggressive environments that may cause their strength and stiffness to decrease during the plant service life. Among the factors recognized as having the potential to cause structural deterioration are uniform, pitting or crevice corrosion; fatigue, including crack initiation and propagation to fracture; elevated temperature; and irradiation. The evaluation of steel containments and liners for continued service must provide assurance that they are able to withstand future extreme loads during the service period with a level of reliability that is sufficient for public safety. Rational methodologies to provide such assurances can be developed using modern structural reliability analysis principles that take uncertainties in loading, strength, and degradation resulting from environmental factors into account. The research described in this report is in support of the Steel Containments and Liners Program being conducted for the US Nuclear Regulatory Commission by the Oak Ridge National Laboratory. The research demonstrates the feasibility of using reliability analysis as a tool for performing condition assessments and service life predictions of steel containments and liners. Mathematical models that describe time-dependent changes in steel due to aggressive environmental factors are identified, and statistical data supporting the use of these models in time-dependent reliability analysis are summarized. The analysis of steel containment fragility is described, and simple illustrations of the impact on reliability of structural degradation are provided. The role of nondestructive evaluation in time-dependent reliability analysis, both in terms of defect detection and sizing, is examined. A Markov model provides a tool for accounting for time-dependent changes in damage condition of a structural component or system. 151 refs

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

Fusion blanket inherent safety assessment

International Nuclear Information System (INIS)

Sze, D.K.; Jung, J.; Cheng, E.T.

1986-01-01

Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

Mathematical Safety Assessment Approaches for Thermal Power Plants

Directory of Open Access Journals (Sweden)

Zong-Xiao Yang

2014-01-01

Full Text Available How to use system analysis methods to identify the hazards in the industrialized process, working environment, and production management for complex industrial processes, such as thermal power plants, is one of the challenges in the systems engineering. A mathematical system safety assessment model is proposed for thermal power plants in this paper by integrating fuzzy analytical hierarchy process, set pair analysis, and system functionality analysis. In the basis of those, the key factors influencing the thermal power plant safety are analyzed. The influence factors are determined based on fuzzy analytical hierarchy process. The connection degree among the factors is obtained by set pair analysis. The system safety preponderant function is constructed through system functionality analysis for inherence properties and nonlinear influence. The decision analysis system is developed by using active server page technology, web resource integration, and cross-platform capabilities for applications to the industrialized process. The availability of proposed safety assessment approach is verified by using an actual thermal power plant, which has improved the enforceability and predictability in enterprise safety assessment.

Towards harmonised self assessment of research reactor safety status in operating organisations

International Nuclear Information System (INIS)

Kirchsteiger, C.; Boeck, H.

2006-01-01

The objective of this paper is to describe the development of a methodology and corresponding web-based tool for mapping and cross-comparing the safety approaches in European and other Research Reactor (RR) facilities in order to detect the principal similarities and differences. As an example, the performance of a Probabilistic Safety Assessment (PSA) for RRs is mapped, as follows: is PSA performed at all? (Yes/No); if so, is PSA mandatory or just recommended? (Yes/No); what is the scope of PSA?, its objective? and practical use? (set of more detailed questions), etc. In this way, information on different types of safety verification practices and requirements for RRs from Europe, Argentina, Australia, Canada, South Africa and the USA has been collected in a systematic way and included in the web-based benchmarking tool DARES (DAtabase for REsearch Reactor Safety). DARES has been developed and filled with sample data by the European Commission's Joint Research Centre (JRC) together with members of the European Research Reactors Operator Group (RROG). A systematic mapping by using DARES in parallel to an international Working Group, consisting of both operators and authorities could be the starting point towards harmonisation of RR safety verification on an international level. In addition, the availability of a user-friendly Information System on the Internet such as DARES containing this information is considered a useful mechanism to exchange international experiences and practices in the area among qualified users. This approach is currently considered to be proposed to the International Atomic Energy Agency (IAES) as one possible application of the recently adopted IAEA Code of Conduct on the Safety of Research Reactors. The resulting process would be a self-assessment of the RR safety status in regulatory bodies and operating organisations relative to the guidance in the Code, practically realised and monitored by an Information System similar to DARES. (orig.)

Buffer and backfill process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrik (comp.)

2006-09-15

This document compiles information on processes in the buffer and deposition tunnel backfill relevant for long-term safety of a KBS-repository. It supports the safety assessment SR-Can, which is a preparatory step for a safety assessment that will support the licence application for a final repository in Sweden. The purpose of the process reports is to document the scientific knowledge of the processes to a level required for an adequate treatment of the processes in the safety assessment. The documentation is not exhaustive from a scientific point of view, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. However, it must be sufficiently detailed to motivate, by arguments founded on scientific understanding, the treatment of each process in the safety assessment. The purpose is further to determine how to handle each process in the safety assessment at an appropriate degree of detail, and to demonstrate how uncertainties are taken care of, given the suggested handling.

Buffer and backfill process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Sellin, Patrik

2006-09-01

This document compiles information on processes in the buffer and deposition tunnel backfill relevant for long-term safety of a KBS-repository. It supports the safety assessment SR-Can, which is a preparatory step for a safety assessment that will support the licence application for a final repository in Sweden. The purpose of the process reports is to document the scientific knowledge of the processes to a level required for an adequate treatment of the processes in the safety assessment. The documentation is not exhaustive from a scientific point of view, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. However, it must be sufficiently detailed to motivate, by arguments founded on scientific understanding, the treatment of each process in the safety assessment. The purpose is further to determine how to handle each process in the safety assessment at an appropriate degree of detail, and to demonstrate how uncertainties are taken care of, given the suggested handling

Food safety: importance of composition for assessing genetically modified cassava (Manihot esculenta Crantz).

Science.gov (United States)

van Rijssen, Fredrika W Jansen; Morris, E Jane; Eloff, Jacobus N

2013-09-04

The importance of food composition in safety assessments of genetically modified (GM) food is described for cassava ( Manihot esculenta Crantz) that naturally contains significantly high levels of cyanogenic glycoside (CG) toxicants in roots and leaves. The assessment of the safety of GM cassava would logically require comparison with a non-GM crop with a proven "history of safe use". This study investigates this statement for cassava. A non-GM comparator that qualifies would be a processed product with CG level below the approved maximum level in food and that also satisfies a "worst case" of total dietary consumption. Although acute and chronic toxicity benchmark CG values for humans have been determined, intake data are scarce. Therefore, the non-GM cassava comparator is defined on the "best available knowledge". We consider nutritional values for cassava and conclude that CG residues in food should be a priority topic for research.

A Computer Program for Assessing Nuclear Safety Culture Impact

Energy Technology Data Exchange (ETDEWEB)

Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

2014-10-15

Through several accidents of NPP including the Fukushima Daiichi in 2011 and Chernobyl accidents in 1986, a lack of safety culture was pointed out as one of the root cause of these accidents. Due to its latent influences on safety performance, safety culture has become an important issue in safety researches. Most of the researches describe how to evaluate the state of the safety culture of the organization. However, they did not include a possibility that the accident occurs due to the lack of safety culture. Because of that, a methodology for evaluating the impact of the safety culture on NPP's safety is required. In this study, the methodology for assessing safety culture impact is suggested and a computer program is developed for its application. SCII model which is the new methodology for assessing safety culture impact quantitatively by using PSA model. The computer program is developed for its application. This program visualizes the SCIs and the SCIIs. It might contribute to comparing the level of the safety culture among NPPs as well as improving the management safety of NPP.

Healthcare professionals? views on feedback of a patient safety culture assessment

OpenAIRE

Zwijnenberg, Nicolien C.; Hendriks, Michelle; Hoogervorst-Schilp, Janneke; Wagner, Cordula

2016-01-01

Background By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals? views on the feedback of a patient safety culture assessment. Methods Twenty four hospitals participated in a patient safety culture assessment in 2012. Hospital departments received feedback in a report and on a websi...

Safety assessment of the liquid-fed ceramic melter process

International Nuclear Information System (INIS)

Buelt, J.L.; Partain, W.L.

1980-08-01

As part of its development program for the solidification of high-level nuclear waste, Pacific Northwest Laboratory assessed the safety issues for a complete liquid-fed ceramic melter (LFCM) process. The LFCM process, an adaption of commercial glass-making technology, is being developed to convert high-level liquid waste from the nuclear fuel cycle into glass. This safety assessment uncovered no unresolved or significant safety problems with the LFCM process. Although in this assessment the LFCM process was not directly compared with other solidification processes, the safety hazards of the LFCM process are comparable to those of other processes. The high processing temperatures of the glass in the LFCM pose no additional significant safety concerns, and the dispersible inventory of dried waste (calcine) is small. This safety assessment was based on the nuclear power waste flowsheet, since power waste is more radioactive than defense waste at the time of solidification, and all accident conditions for the power waste would have greater radiological consequences than those for defense waste. An exhaustive list of possible off-standard conditions and equipment failures was compiled. These accidents were then classified according to severity of consequence and type of accident. Radionuclide releases to the stack were calculated for each group of accidents using conservative assumptions regarding the retention and decontamination features of the process and facility. Two recommendations that should be considered by process designers are given in the safety assessment

Performance-based containment leakage testing

International Nuclear Information System (INIS)

Cybulskis, P.

1995-01-01

The U.S. Nuclear Regulatory Commission (NRC) is reviewing regulatory requirements in an effort to revise those that are marginal to safety but impose significant burdens on licensees. Identification of requirements marginal to safety and development and evaluation of alternatives utilize the NRC safety goals and insights from probabilistic risk assessments (PRAs). Since earlier studies found design-basis containment leakage to be a minor contributor to reactor accident risk, containment leakage testing has been selected as a candidate for change in regulations. This paper summarizes the technical analyses supporting the NRC proposal to amend Appendix J of 10 CFR Part 50 as its first effort to decrease unnecessary regulatory burdens on licensees

Safety assessment of novel foods and strategies to determine their safety in use

International Nuclear Information System (INIS)

Edwards, Gareth

2005-01-01

Safety assessment of novel foods requires a different approach to that traditionally used for the assessment of food chemicals. A case-by-case approach is needed which must be adapted to take account of the characteristics of the individual novel food. A thorough appraisal is required of the origin, production, compositional analysis, nutritional characteristics, any previous human exposure and the anticipated use of the food. The information should be compared with a traditional counterpart of the food if this is available. In some cases, a conclusion about the safety of the food may be reached on the basis of this information alone, whereas in other cases, it will help to identify any nutritional or toxicological testing that may be required to further investigate the safety of the food. The importance of nutritional evaluation cannot be over-emphasised. This is essential for the conduct of toxicological studies in order to avoid dietary imbalances, etc., that might lead to interpretation difficulties, but also in the context of its use as food and to assess the potential impact of the novel food on the human diet. The traditional approach used for chemicals, whereby an acceptable daily intake (ADI) is established with a large safety margin relative to the expected exposure, cannot be applied to foods. The assessment of safety in use should be based upon a thorough knowledge of the composition of the food, evidence from nutritional, toxicological and human studies, expected use of the food and its expected consumption. Safety equates to a reasonable certainty that no harm will result from intended uses under the anticipated conditions of consumption

LANL Safety Conscious Work Environment (SCWE) Self-Assessment

Energy Technology Data Exchange (ETDEWEB)

Hargis, Barbara C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-01-29

On December 21, 2012 Secretary of Energy Chu transmitted to the Defense Nuclear Facilities Safety Board (DNFSB) revised commitments on the implementation plan for Safety Culture at the Waste Treatment and Immobilization Plant. Action 2-5 was revised to require contractors and federal organizations to complete Safety Conscious Work Environment (SCWE) selfassessments and provide reports to the appropriate U.S. Department of Energy (DOE) - Headquarters Program Office by September 2013. Los Alamos National Laboratory (LANL) planned and conducted a Safety Conscious Work Environment (SCWE) Self-Assessment over the time period July through August, 2013 in accordance with the SCWE Self-Assessment Guidance provided by DOE. Significant field work was conducted over the 2-week period August 5-16, 2013. The purpose of the self-assessment was to evaluate whether programs and processes associated with a SCWE are in place and whether they are effective in supporting and promoting a SCWE.

Quantitative risk assessment of digitalized safety systems

Energy Technology Data Exchange (ETDEWEB)

Shin, Sung Min; Lee, Sang Hun; Kang, Hym Gook [KAIST, Daejeon (Korea, Republic of); Lee, Seung Jun [UNIST, Ulasn (Korea, Republic of)

2016-05-15

A report published by the U.S. National Research Council indicates that appropriate methods for assessing reliability are key to establishing the acceptability of digital instrumentation and control (I and C) systems in safety-critical plants such as NPPs. Since the release of this issue, the methodology for the probabilistic safety assessment (PSA) of digital I and C systems has been studied. However, there is still no widely accepted method. Kang and Sung found three critical factors for safety assessment of digital systems: detection coverage of fault-tolerant techniques, software reliability quantification, and network communication risk. In reality the various factors composing digitalized I and C systems are not independent of each other but rather closely connected. Thus, from a macro point of view, a method that can integrate risk factors with different characteristics needs to be considered together with the micro approaches to address the challenges facing each factor.

Survey of probabilistic methods in safety and risk assessment for nuclear power plant licensing

International Nuclear Information System (INIS)

1984-04-01

After an overview about the goals and general methods of probabilistic approaches in nuclear safety the main features of probabilistic safety or risk assessment (PRA) methods are discussed. Mostly in practical applications not a full-fledged PRA is applied but rather various levels of analysis leading from unavailability assessment of systems over the more complex analysis of the probable core damage stages up to the assessment of the overall health effects on the total population from a certain practice. The various types of application are discussed in relation to their limitation and benefits for different stages of design or operation of nuclear power plants. This gives guidance for licensing staff to judge the usefulness of the various methods for their licensing decisions. Examples of the application of probabilistic methods in several countries are given. Two appendices on reliability analysis and on containment and consequence analysis provide some more details on these subjects. (author)

Safety assessment of high consequence robotics system

International Nuclear Information System (INIS)

Robinson, D.G.; Atcitty, C.B.

1996-01-01

This paper outlines the use of a failure modes and effects analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, the weigh and leak check system, is to replace a manual process for weight and leakage of nuclear materials at the DOE Pantex facility. Failure modes and effects analyses were completed for the robotics process to ensure that safety goals for the systems have been met. Due to the flexible nature of the robot configuration, traditional failure modes and effects analysis (FMEA) were not applicable. In addition, the primary focus of safety assessments of robotics systems has been the protection of personnel in the immediate area. In this application, the safety analysis must account for the sensitivities of the payload as well as traditional issues. A unique variation on the classical FMEA was developed that permits an organized and quite effective tool to be used to assure that safety was adequately considered during the development of the robotic system. The fundamental aspects of the approach are outlined in the paper

Liquefied gaseous fuels safety and environmental control assessment program: third status report

Energy Technology Data Exchange (ETDEWEB)

1982-03-01

This Status Report contains contributions from all contractors currently participating in the DOE Liquefied Gaseous Fuels (LG) Safety and Environmental Control Assessment Program and is presented in two principal sections. Section I is an Executive Summary of work done by all program participants. Section II is a presentation of fourteen individual reports (A through N) on specific LGF Program activities. The emphasis of Section II is on research conducted by Lawrence Livermore National Laboratory (Reports A through M). Report N, an annotated bibliography of literature related to LNG safety and environmental control, was prepared by Pacific Northwest Laboratory (PNL) as part of its LGF Safety Studies Project. Other organizations who contributed to this Status Report are Aerojet Energy Conversion Company; Applied Technology Corporation; Arthur D. Little, Incorporated; C/sub v/ International, Incorporated; Institute of Gas Technology; and Massachusetts Institute of Technology. Separate abstracts have been prepared for Reports A through N for inclusion in the Energy Data Base.

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

Research on fuzzy comprehensive assessment method of nuclear power plant safety culture

International Nuclear Information System (INIS)

Xiang Yuanyuan; Chen Xukun; Xu Rongbin

2012-01-01

Considering the traits of safety culture in nuclear plant, 38 safety culture assessment indexes are established from 4 aspects such as safety values, safety institution, safety behavior and safety sub- stances. Based on it, a comprehensive assessment method for nuclear power plant safety culture is constructed by using AHP (Analytic Hierarchy Process) approach and fuzzy mathematics. The comprehensive assessment method has the quality of high precision and high operability, which can support the decision making of safety culture development. (authors)

Safety/security interface assessments at commercial nuclear power plants

International Nuclear Information System (INIS)

Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

1985-01-01

The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur

Safety/security interface assessments at commercial nuclear power plants

International Nuclear Information System (INIS)

Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

1985-07-01

The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur. 2 refs

Safety assessment for Generation IV nuclear systems

International Nuclear Information System (INIS)

Leahy, T.J.

2012-01-01

The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Recent RSWG work has focused on the definition of an integrated safety assessment methodology (ISAM) for evaluating the safety of Generation IV systems. ISAM is an integrated 'tool-kit' consisting of 5 analytical techniques that are available and matched to appropriate stages of Generation IV system concept development: 1) qualitative safety features review - QSR, 2) phenomena identification and ranking table - PIRT, 3) objective provision tree - OPT, 4) deterministic and phenomenological analyses - DPA, and 5) probabilistic safety analysis - PSA. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time

Nirex safety assessment research programme: 1987/88

International Nuclear Information System (INIS)

George, D.; Hodgkinson, D.P.

1987-01-01

The Nirex Safety Assessment Research programme's objective is to provide information for the radiological safety case for disposing low-level and intermediate-level radioactive wastes in underground repositories. The programme covers a wide range of experimental studies and mathematical modelling for the near and far field. It attempts to develop a quantitative understanding of events and processes which have an impact on the safety of radioactive waste disposal. (U.K.)

Nuclear safety and regulation

International Nuclear Information System (INIS)

Kim, Hho Jung

2000-03-01

This book contains 12 chapters, which are atom and radiation, nuclear reactor and kinds of nuclear power plant, safeguard actuation system and stability evaluation for rock foundation of nuclear power plant, nuclear safety and principle, safety analysis and classification of incident, probabilistic safety assessment and major incident, nuclear safety regulation, system of nuclear safety regulation, main function and subject of safety regulation in nuclear facilities, regulation of fuel cycle and a nuclear dump site, protection of radiation and, safety supervision and, safety supervision and measurement of environmental radioactivity.

Geosphere process report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Skagius, Kristina

2010-11-01

This report documents geosphere processes identified as relevant to the long-term safety of a KBS-3 repository, and forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

Geosphere process report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Skagius, Kristina (ed.) (Kemakta Konsult AB, Stockholm (Sweden))

2010-11-15

This report documents geosphere processes identified as relevant to the long-term safety of a KBS-3 repository, and forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

Research on advanced system safety assessment procedures (4)

International Nuclear Information System (INIS)

Suzuki, Kazuhiko; Shimada, Yukiyasu

2001-03-01

The past research reports in the area of safety engineering proposed the Computer-aided HAZOP system to be applied to Nuclear Reprocessing Facilities. Automated HAZOP system has great advantage compared with human analysts in terms of accuracy of the results, and time required to conduct HAZOP studies. This report surveys the literature on risk assessment and safety design based on the concept of independent protection layers (IPLs). Furthermore, to improve HAZOP System, tool is proposed to construct the basic model and the internal state model. Such HAZOP system is applied to analyze two kinds of processes, where the ability of the proposed system is verified. In addition, risk assessment support system is proposed to integrate safety design environment and assessment result to be used by other plants as well as to enable the underline plant to use other plants' information. This technique can be implemented using web-based safety information systems. (author)

Audit of data and code use in the SR-Can safety assessment

Energy Technology Data Exchange (ETDEWEB)

Hicks, T.W.; Baldwin, T.D. [Galson Sciences Ltd, 5 Grosvenor House, Melton R oad, Oakham, Rutland LE15 6AX (United Kingdom)

2008-03-15

Building on the findings of previous studies on data and code quality assurance (QA) in safety assessments, this report provides a review of data and code QA in the SR-Can safety assessment. The data quality audit aimed to check that the selection and use of data in the SR-Can safety assessment was appropriate, focusing on the data that underpin representations of and assumptions about canister, insert, buffer, and backfill behaviour. The SR-Can Data Report provided the initial focus for examining the traceability and reliability of data used in the safety assessment; the Data Report is one of the series of SR-Can safety assessment reports and, in this review, it was anticipated that it would provide the primary source of data on the canister, insert, buffer, and backfill. However, other safety assessment reports (the SR-Can Main Report, the Initial State Report, the Fuel and Canister Process Report, and the Buffer and Backfill Process Report) were found to provide key information on data used in the safety assessment. The quality audit of codes aimed to check that code use in the SR-Can safety assessment has been justified through a transparent and traceable process of code development and selection. The Model Summary Report provided the focus for reviewing the QA status of the codes used in the safety assessment. As well as highlighting a number of concerns regarding QA aspects of specific data sets, parameter values, and codes used in the SR-Can safety assessment (which are presented in the report), the review has led to several general observations on data and code QA that should be considered by SKB in the development and implementation of a QA system for the SR-Site safety assessment: - The SR-Site safety assessment and associated QA records should include information that demonstrates that a full QA system has been implemented in order to build confidence in the validity of the assessment. - The data and parameter values used directly in the safety

Assessing safety culture using RADAR matrix

International Nuclear Information System (INIS)

Mariscal-Saldana, M. a.; Garcia-Herrero, S.; Toca-Otero, A.

2009-01-01

Santa Maria de Garona nuclear power plant, in collaboration with Burgos University, has proceeded to conduct a pilot project aimed at seeing the possibilities for the RADAR (Results, Approach, Development, Assessment and review) logic of EFQM model, as a tool for self evaluation of Safety Culture in a nuclear power plant. In the work it has sought evidences of Safety culture implanted in the plant, and identify strengths and areas for improvement regarding this Culture. the score obtained by analyzing these strengths and areas for improvements has served to prioritize actions implemented. The nuclear power plant has been submitted voluntarily to the mission SCART (Safety Culture Assessment Review Team), an international review being done for the first time in the world at a plant in operation and the team of experts led by International Agency of Atomic Energy (IAEA) has identified this project as a good practice, an innovative process implemented in the plant, that must be transmitted to other plants. (Author) 10 refs

Developing IAM for Life Cycle Safety Assessment

NARCIS (Netherlands)

Toxopeus, Marten E.; Lutters, Diederick; Nee, Andrew Y.C.; Song, Bin; Ong, Soh-Khim

2013-01-01

This publication discusses aspects of the development of an impact assessment method (IAM) for safety. Compared to the many existing IAM’s for environmentally oriented LCA, this method should translate the impact of a product life cycle on the subject of safety. Moreover, the method should be

Safety assessment of smoke flavouring primary products by the European Food Safety Authority

NARCIS (Netherlands)

Theobald, A.; Arcella, D.; Carere, A.; Croera, C.; Engel, K.H.; Gott, D.; Gurtler, R.; Meier, D.; Pratt, I.; Rietjens, I.M.C.M.; Simon, R.; Walker, R.

2012-01-01

This paper summarises the safety assessments of eleven smoke flavouring primary products evaluated by the European Food Safety Authority (EFSA). Data on chemical composition, content of polyaromatic hydrocarbons and results of genotoxicity tests and subchronic toxicity studies are presented and

Technical findings and regulatory analysis for Generic Safety Issue II.E.4.3, ''Containment Integrity Check''

International Nuclear Information System (INIS)

Serkiz, A.W.

1988-04-01

This report contains the technical findings and regulatory analysis for Generic Safety Issue II.E.4.3, ''Containment Integrity Check.'' An evaluation of the containment isolation history from 1965 to 1983 reveals that (except for a small number of events) containment integrity has been maintained and that the majority of reported events have been events related to exceeding Technical Specification limits (or 0.6 of the allowable leakage level). In addition, more recent risk analyses have shown that allowable leakage rates even if increased by a factor of 10 would not significantly increase risk. Potential methods of continuous monitoring are identified and evaluated. Therefore, these technical findings and risk evaluations support closure of Generic Safety Issue II.E.4.3

Fire safety assessment of tunnel structures

DEFF Research Database (Denmark)

Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

2011-01-01

.g. structural and non structural, organizational, human behavior). This is even more truth for the fire safety design of such structures. Fire safety in tunnels is challenging because of the particular environment, bearing in mind also that a fire can occur in different phases of the tunnel’s lifecycle. Plans...... for upgrading fire safety provisions and tunnel management are also important for existing tunnels. In this study, following a brief introduction of issues regarding the above mentioned aspects, the structural performance of a steel rib for a tunnel infrastructure subject to fire is assessed by means...

Safety case methodology for decommissioning of research reactors. Assessment of the long term impact of a flooding scenario

International Nuclear Information System (INIS)

Vladescu, G.; Banciu, O.

1999-01-01

The paper contains the assessment methodology of a Safety Case fuel decommissioning of research reactors, taking into account the international approach principles. The paper also includes the assessment of a flooding scenario for a decommissioned research reactor (stage 1 of decommissioning). The scenario presents the flooding of reactor basement, radionuclide migration through environment and long term radiological impact for public. (authors)

Safety management system needs assessment.

Science.gov (United States)

2016-04-01

The safety of the traveling public is critical as each year there are approximately 200 highway fatalities in Nebraska and numerous crash injuries. The objective of this research was to conduct a needs assessment to identify the requirements of a sta...

Seismic test for safety evaluation of low level radioactive wastes containers

International Nuclear Information System (INIS)

Ohoka, Makoto; Horikiri, Morito

1998-08-01

Seismic safety of three-piled container system used in Tokai reprocessing center was confirmed by seismic test and computational analysis. Two types of container were evaluated, for low level noninflammable radioactive solid wastes, and for used filters wrapped by large plastic bags. Seismic integrity of three-piled containers was confirmed by evaluating response characteristics such as acceleration and displacement under the design earthquake condition S1, which is the maximum earthquake expected at the stored site during the storage time. Computational dynamic analysis was also performed, and several conclusions described below were made. (1) Response characteristics of the bottom board and the side board were different. The number of pile did not affect the response characteristics of the bottom board of each container. They behaved as a rigid body. (2) The response of the side board was larger than that of the bottom board. (3) The response depended on the direction in each board, either side or bottom. The response acceleration became larger to the seismic wave perpendicular to the plane which has the entrance for fork lift and the radioactive warning mark. (4) The maximum horizontal response displacement under the S1 seismic wave was approximately 10 mm. It is so small that it does not affect the seismic safety. (5) The stoppers to prevent fall down had no influence to the response acceleration. (6) There was no fall down to the S1 seismic wave and 2 times of S1 seismic wave, which was the maximum input condition of the test. (7) The response of the bottom board of the containers, which are main elements of fall down, had good agreements both in the test and in the computational analysis. (author)

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Energy Technology Data Exchange (ETDEWEB)

Lee, Seunghee; Kim, Juyoul, E-mail: gracemi@fnctech.com

2017-03-15

Highlights: • Post-closure safety assessment of near surface disposal facility for DSRS was performed. • Engineered vault and rock-cavern type were considered for normal and well scenario. • {sup 14}C, {sup 226}Ra, {sup 241}Am were primary nuclides contributing large portion of exposure dose. • Near surface disposal of DSRSs containing {sup 14}C, {sup 226}Ra and {sup 241}Am should be restricted. - Abstract: Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1 × 10{sup −7} mSv/yr, for both disposal options and satisfied the regulatory limit

Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

Energy Technology Data Exchange (ETDEWEB)

Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

1995-12-31

The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

Safety and security risk assessments--now demystified!

Science.gov (United States)

White, Donald E

2011-01-01

Safety/security risk assessments no longer need to spook nor baffle healthcare safety/security managers. This grid template provides at-at-glance quick lookup of the possible threats, the affected people and things, a priority ranking of these risks, and a workable solution for each risk. Using the standard document, spreadsheet, or graphics software already available on your computer, you can easily use a scientific method to produce professional looking risk assessments that get quickly understood by both senior managers and first responders alike!

Probabilistic safety and risk assessments in the field of nuclear technology - Mode of operation, possibilities and limits

International Nuclear Information System (INIS)

Mertens, J.

1993-01-01

In this study probabilistic safety and risk assessments in the field of nuclear energy are explained. Mainly qualitative results and conclusions are presented. Explanations for often discussed aspects of such analysis reveal the procedure and reasonable limits of application. The mentioned literature contains detailed results. (orig./DG) [de

Visualization of Safety Assessment Result Using GIS in SITES

International Nuclear Information System (INIS)

Yun, Bong-Yo; Park, Joo Wan; Park, Se-Moon; Kim, Chang-Lak

2006-01-01

Site Information and Total Environmental database management System (SITES) is an integrated program for overall data analysis, environmental monitoring, and safety analysis that are produced from the site investigation and environmental assessment of the relevant nuclear facility. SITES is composed of three main modules such as Site Environment Characterization database for Unified and Reliable Evaluation system (SECURE), Safety Assessment INTegration system (SAINT) and Site Useful Data Analysis and ALarm system (SUDAL). The visualization function of safety assessment and environmental monitoring results is designed. This paper is to introduce the visualization design method using Geographic Information System (GIS) for SITES

Development of internal dose assessment procedure for workers in industries using raw materials containing naturally occurring radioactive materials

International Nuclear Information System (INIS)

Choi, Cheol Kyu; KIm, Yong Geon; Ji, Seung Woo; Kim, Kwang Pyo; Koo, Bon Cheol; Chang, Byung Uck

2016-01-01

It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are 10 Bq·g-1 for 40K and 1 Bq·g-1 for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups (<0.1 mSv, 0.1-0.3 mSv, and >0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels (<0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and >1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries

Development of internal dose assessment procedure for workers in industries using raw materials containing naturally occurring radioactive materials

Energy Technology Data Exchange (ETDEWEB)

Choi, Cheol Kyu; KIm, Yong Geon; Ji, Seung Woo; Kim, Kwang Pyo [College of Engineering, Kyung Hee University, Yongin (Korea, Republic of); Koo, Bon Cheol; Chang, Byung Uck [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-09-15

It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are 10 Bq·g-1 for 40K and 1 Bq·g-1 for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups (<0.1 mSv, 0.1-0.3 mSv, and >0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels (<0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and >1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries.

Containment Emergency Sump Performance. Technical findings related to Unresolved Safety Issue A-43. Revision 1

International Nuclear Information System (INIS)

1985-10-01

This report summarizes key technical findings related to Unresolved Safety Issue (USI) A-43, Containment Emergency Sump Performance. Both BWRs and PWRs are considered in this report. Emergency core cooling systems require a clean, reliable water source to maintain long-term recirculation following a loss-of-coolant accident (LOCA). PWRs rely on the containment emergency sump to provide such a water supply to residual heat removal pumps and containment spray pumps. BWRs rely on pump suction intakes in the suppression pool or wet well to provide water to residual heat removal and core spray systems. Thus, the technical findings in this report provide information on post-LOCA recirculation. These findings have been derived from extensive experimental studies, generic plant studies, and assessments of sumps used for long-term cooling. Results of hydraulic tests have shown that the potential for air ingestion is less severe than previously hypothesized. The effects of debris blockage on NPSH margin must be dealt with on a plant-specific basis. These findings have been used to develop revisions to Regulatory Guide 1.82 and Standard Review Plan Section 6.2.2 (NUREG-0800)

Additional safety assessment of ITER - Addition safety investigation of the INB ITER

International Nuclear Information System (INIS)

2012-01-01

This assessment aims at re-assessing safety margins in the light of events which occurred in Fukushima Daiichi, i.e. extreme natural events challenging the safety of installations. After a presentation of some characteristics of the ITER installation (location, activities, buildings, premise detritiation systems, electric supply, handling means, radioactive materials, chemical products, nuclear risks, specific risks), the report addresses the installation robustness by identifying cliff-edge effect risks which can be related to a loss of confinement of radioactive materials, explosions, a significant increase of exposure level, a possible effect on water sheets, and so on. The next part addresses the various aspects related to a seismic risk: installation sizing (assessment methodology, seismic risk characterization in Cadarache), sizing protection measures, installation compliance, and margin assessment. External flooding is the next addressed risk: installation sizing with respect to this specific risk, protection measures, installation compliance, margin assessment, and studied additional measures. Other extreme natural phenomena are considered (meteorological conditions, earthquake and flood) which may have effects on other installations (dam, canal). Then, the report addresses technical risks like the loss of electric supplies and cooling systems, the way a crisis is managed in terms of technical and human means and organization in different typical accidental cases. Subcontracting practices are also discussed. A synthesis proposes an overview of this additional safety assessment and discusses the impact which could have additional measures which could be implemented

Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

International Nuclear Information System (INIS)

Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

1984-01-01

The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

A regulatory view of containment integrity in the United Kingdom

International Nuclear Information System (INIS)

Bradford, P.M.; Patchett, C.M.

1994-01-01

This paper reviews the approach of HM Nuclear Installations Inspectorate (NII) to containment integrity in the United Kingdom (UK). NII is that part of the regulatory authority, the Health and Safety Executive (HSE), which administers the UK's nuclear site licensing system. A major part of the licensing process lies in the assessment of licensees' submissions for new and existing plant. The purpose of this paper is to: briefly review our revised Safety Assessment Principles, describe our assessment and inspection activities on the primary containment building of the Sizewell B PWR which is progressing to full power operation in 1994 and, to indicate our views on the possible directions for future research into containment design and performance. (author). 5 refs

Research on the improvement of nuclear safety

International Nuclear Information System (INIS)

Yoo, Keon Joong; Kim, Dong Soo; Kim, Hui Dong; Park, Chang Kyu

1993-06-01

To improve the nuclear safety, this project is divided into three areas which are the development of safety analysis technology, the development of severe accident analysis technology and the development of integrated safety assessment technology. 1. The development of safety analysis technology. The present research aims at the development of necessary technologies for nuclear safety analysis in Korea. Establishment of the safety analysis technologies enables to reduce the expenditure both by eliminating excessive conservatisms incorporated in nuclear reactor design and by increasing safety margins in operation. It also contributes to improving plant safety through realistic analyses of the Emergency Operating Procedures (EOP). 2. The development of severe accident analysis technology. By the computer codes (MELCOR and CONTAIN), the in-vessel and the ex-vessel severe accident phenomena are simulated. 3. The development of integrated safety assessment technology. In the development of integrated safety assessment techniques, the included research areas are the improvement of PSA computer codes, the basic study on the methodology for human reliability analysis (HRA) and common cause failure (CCF). For the development of the level 2 PSA computer code, the basic research for the interface between level 1 and 2 PSA, the methodology for the treatment of containment event tree are performed. Also the new technologies such as artificial intelligence, object-oriented programming techniques are used for the improvement of computer code and the assessment techniques

Assessment of airborne asbestos exposure during the servicing and handling of automobile asbestos-containing gaskets.

Science.gov (United States)

Blake, Charles L; Dotson, G Scott; Harbison, Raymond D

2006-07-01

Five test sessions were conducted to assess asbestos exposure during the removal or installation of asbestos-containing gaskets on vehicles. All testing took place within an operative automotive repair facility involving passenger cars and a pickup truck ranging in vintage from late 1960s through 1970s. A professional mechanic performed all shop work including engine disassembly and reassembly, gasket manipulation and parts cleaning. Bulk sample analysis of removed gaskets through polarized light microscopy (PLM) revealed asbestos fiber concentrations ranging between 0 and 75%. Personal and area air samples were collected and analyzed using National Institute of Occupational Safety Health (NIOSH) methods 7400 [phase contrast microscopy (PCM)] and 7402 [transmission electron microscopy (TEM)]. Among all air samples collected, approximately 21% (n = 11) contained chrysotile fibers. The mean PCM and phase contrast microscopy equivalent (PCME) 8-h time weighted average (TWA) concentrations for these samples were 0.0031 fibers/cubic centimeters (f/cc) and 0.0017 f/cc, respectively. Based on these findings, automobile mechanics who worked with asbestos-containing gaskets may have been exposed to concentrations of airborne asbestos concentrations approximately 100 times lower than the current Occupational Safety and Health Administration (OSHA) Permissible Exposure Limit (PEL) of 0.1 f/cc.

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

Complementary assessment of the safety of French nuclear power plants

International Nuclear Information System (INIS)

Camarcat, N.; Pouget-Abadie, X.

2011-01-01

As an immediate consequence of the Fukushima accident the French nuclear safety Authority (ASN) asked EDF to perform a complementary safety assessment for each nuclear power plant dealing with 3 points: 1) the consequences of exceptional natural disasters, 2) the consequences of total loss of electrical power, and 3) the management of emergency situations. The safety margin has to be assessed considering 3 main points: first a review of the conformity to the initial safety requirements, secondly the resistance to events overdoing what the facility was designed to stand for, and the feasibility of any modification susceptible to improve the safety of the facility. This article details the specifications of such assessment, the methodology followed by EDF, the task organization and the time schedule. (A.C.)

Assessment of safety regulation using an artificial society

International Nuclear Information System (INIS)

Furuta, Kazuo; Nagase, Masaya

2005-01-01

This study proposes using an artificial society to assess impacts of safety regulation on the society. The artificial society used in this study is a multi-agent system, which consists of many agents representing companies. The agents cannot survive unless they get profits by producing some products. Safety regulation functions as the business environment, which the agents will evolve to fit to. We modeled this process of survival and adaptation by the genetic algorithm. Using the proposed model, case simulations were performed to compare various regulation styles, and some interesting insights were obtained how regulation style influences behavior of the agents and then productivity and safety level of the industry. In conclusion, an effective method for assessment of safety regulation has been developed, and then several insights were shown in this study

Integrated Deterministic-Probabilistic Safety Assessment Methodologies

Energy Technology Data Exchange (ETDEWEB)

Kudinov, P.; Vorobyev, Y.; Sanchez-Perea, M.; Queral, C.; Jimenez Varas, G.; Rebollo, M. J.; Mena, L.; Gomez-Magin, J.

2014-02-01

IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be based on the coupling of deterministic (consequences) and probabilistic (frequency) considerations to address the mutual interactions between stochastic disturbances (e.g. failures of the equipment, human actions, stochastic physical phenomena) and deterministic response of the plant (i.e. transients). This paper gives a general overview of some IDPSA methods as well as some possible applications to PWR safety analyses. (Author)

Human reliability in probabilistic safety assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

Nowadays a growing interest in medioambiental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processess and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects. (This relevance has been demostrated in the accidents happenned). However in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a guide to carry out a Human Reliability Analysis and c) a selected overwiev of the techniques and methodologies currently applied in this area. (Author)

Preclosure radiological safety assessment for the ground support system in the exploratory studies facility

International Nuclear Information System (INIS)

Smith, A.J.; Tsai, F.C.

1995-01-01

An initial probabilistic safety assessment was performed for the exploratory studies facility underground opening to determine whether the ground support system should be classified as an item important to safety. The initiating event was taken to be a rock fall in an operational facility impacting a loaded waste transporter. Rock fall probability rates were estimated from data reported by commercial mining operations. This information was retrieved from the data base compiled by the Mining Safety and Health Administration from the mandatory reporting of incidents. The statistical distribution of the rock fall magnitude was estimated from the horizontal and vertical spacing fractures measured at the Yucca Mountain repository horizon. Simple models were developed to estimate container deformation and radionuclide releases arising from the projected distribution of impacts. Accepted techniques were used to calculate atmospheric dispersion and obtain the committed dose to individuals

Probabilistic assessment of NPP safety under aircraft impact

International Nuclear Information System (INIS)

Birbraer, A.N.; Roleder, A.J.; Arhipov, S.B.

1999-01-01

Methodology of probabilistic assessment of NPP safety under aircraft impact is described below. The assessment is made taking into account not only the fact of aircraft fall onto the NPP building, but another casual parameters too, namely an aircraft class, velocity and mass, as well as point and angle of its impact with the building structure. This analysis can permit to justify the decrease of the required structure strength and dynamic loads on the NPP equipment. It can also be especially useful when assessing the safety of existing NPP. (author)

Nirex Safety Assessment Research Programme bibliography, 1990

International Nuclear Information System (INIS)

Cooper, M.J.

1990-10-01

This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. (author)

Understanding and assessing safety culture

International Nuclear Information System (INIS)

Dalling, Ian

1997-01-01

The 'Dalling' integrated model of organisational performance is introduced and described. A principal element of this model is culture, which is dynamically contrasted with the five other interacting critical elements, which comprise: the management system, the knowledge base, corporate leadership, stakeholders and consciousness. All six of these principal driving elements significantly influence health, safety, environmental, security, or any other aspect of organisational performance. It is asserted that the elements of organisational performance must be clearly defined and understood if meaningful measurements are to be carried out and sustained progress made in improving the knowledge of organisational performance. AEA Technology's safety culture research programme is then described together with the application of a safety culture assessment tool to organisations in the nuclear, electricity, transport, and oil and gas industries, both within and outside of the United Kingdom. (author)

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

International Nuclear Information System (INIS)

Birkhofer, A.

1992-01-01

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

Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

International Nuclear Information System (INIS)

Lee, Sang Hun; Son, Kwang Seop; Jung, Wondea; Kang, Hyun Gook

2017-01-01

Highlights: • Safety data communication risk assessment framework and quantitative scheme were proposed. • Fault-tree model of ESFAS unavailability due to safety data communication failure was developed. • Safety data link and network risk were assessed based on various ESF-CCS design specifications. • The effect of fault-tolerant algorithm reliability of safety data network on ESFAS unavailability was assessed. - Abstract: As one of the safety-critical systems in nuclear power plants (NPPs), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for the transmission of safety component actuation signals from the group controllers to loop controllers to effectively accommodate various safety-critical field controllers. Since data communication failure risk in the ESF-CCS has yet to be fully quantified, the ESF-CCS employing data communication systems have not been applied in NPPs. This study therefore developed a fault tree model to assess the data link and data network failure-induced unavailability of a system function used to generate an automated control signal for accident mitigation equipment. The current aim is to provide risk information regarding data communication failure in a digital safety feature control system in consideration of interconnection between controllers and the fault-tolerant algorithm implemented in the target system. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to data link and network failure and its risk effect on safety signal generation failure. This study is expected to provide insight into the risk assessment of safety-critical data communication in a digitalized NPP instrumentation and control system.

Tooth bleaching using peroxide-containing agents: current status of safety issues.

Science.gov (United States)

Li, Y

1998-08-01

During the last 10 years, at-home tooth bleaching using peroxide-containing agents has quickly become well accepted. It is one of the most popular cosmetic dental procedures for whitening teeth. Although there are few disputes regarding their efficacy, concerns and debates have continued regarding the safety of peroxide-containing tooth bleaching agents. Potential carcinogenicity and genotoxicity of the peroxides used in bleaching agents are the two most persistent and controversial issues. This article reviews and discusses available information on carcinogenicity and genotoxicity of hydrogen peroxide and the potential risks associated with the use of peroxide-containing bleaching agents. Clinical studies reported after the 1996 international symposium on responses of oral tissues to peroxide-containing bleaching agents are also reviewed. Overall evidence supports the conclusion that the proper use of peroxide-containing at-home tooth bleaching agents is safe. However, potential adverse effects may occur in inappropriate applications, abuses, or the use of inappropriate products. At-home tooth bleaching should be monitored by dental professionals to maximize the benefits while minimizing potential risks.

Savannah River Site K-Reactor Probabilistic Safety Assessment

International Nuclear Information System (INIS)

Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

1992-12-01

This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

A novel safety assessment strategy applied to non-selective extracts.

Science.gov (United States)

Koster, Sander; Leeman, Winfried; Verheij, Elwin; Dutman, Ellen; van Stee, Leo; Nielsen, Lene Munch; Ronsmans, Stefan; Noteborn, Hub; Krul, Lisette

2015-06-01

A main challenge in food safety research is to demonstrate that processing of foodstuffs does not lead to the formation of substances for which the safety upon consumption might be questioned. This is especially so since food is a complex matrix in which the analytical detection of substances, and consequent risk assessment thereof, is difficult to determine. Here, a pragmatic novel safety assessment strategy is applied to the production of non-selective extracts (NSEs), used for different purposes in food such as for colouring purposes, which are complex food mixtures prepared from reference juices. The Complex Mixture Safety Assessment Strategy (CoMSAS) is an exposure driven approach enabling to efficiently assess the safety of the NSE by focussing on newly formed substances or substances that may increase in exposure during the processing of the NSE. CoMSAS enables to distinguish toxicologically relevant from toxicologically less relevant substances, when related to their respective levels of exposure. This will reduce the amount of work needed for identification, characterisation and safety assessment of unknown substances detected at low concentration, without the need for toxicity testing using animal studies. In this paper, the CoMSAS approach has been applied for elderberry and pumpkin NSEs used for food colouring purposes. Copyright © 2015 Elsevier Ltd. All rights reserved.

Suggestions on the Development of Safety Culture Assessment Method

International Nuclear Information System (INIS)

Choi, Young Sung; Choi, Kwang Sik; Kim, Woong Sik

2006-01-01

Several efforts have been made to assess safety culture of organization that operates nuclear power plants in Korea. The MOST and KINS played a major role to develop assessment methods and KHNP applied them to its NPPs. This paper explains the two methods developed by KINS briefly and presents the insights obtained from the two different applications. It concludes with some suggestions for safety culture assessment based on the insights

Integrating bioassays and analytical chemistry as an improved approach to support safety assessment of food contact materials.

Science.gov (United States)

Veyrand, Julien; Marin-Kuan, Maricel; Bezencon, Claudine; Frank, Nancy; Guérin, Violaine; Koster, Sander; Latado, Hélia; Mollergues, Julie; Patin, Amaury; Piguet, Dominique; Serrant, Patrick; Varela, Jesus; Schilter, Benoît

2017-10-01

Food contact materials (FCM) contain chemicals which can migrate into food and result in human exposure. Although it is mandatory to ensure that migration does not endanger human health, there is still no consensus on how to pragmatically assess the safety of FCM since traditional approaches would require extensive toxicological and analytical testing which are expensive and time consuming. Recently, the combination of bioassays, analytical chemistry and risk assessment has been promoted as a new paradigm to identify toxicologically relevant molecules and address safety issues. However, there has been debate on the actual value of bioassays in that framework. In the present work, a FCM anticipated to release the endocrine active chemical 4-nonyphenol (4NP) was used as a model. In a migration study, the leaching of 4NP was confirmed by LC-MS/MS and GC-MS. This was correlated with an increase in both estrogenic and anti-androgenic activities as measured with bioassays. A standard risk assessment indicated that according to the food intake scenario applied, the level of 4NP measured was lower, close or slightly above the acceptable daily intake. Altogether these results show that bioassays could reveal the presence of an endocrine active chemical in a real-case FCM migration study. The levels reported were relevant for safety assessment. In addition, this work also highlighted that bioactivity measured in migrate does not necessarily represent a safety issue. In conclusion, together with analytics, bioassays contribute to identify toxicologically relevant molecules leaching from FCM and enable improved safety assessment.

A 90-day safety study in Sprague-Dawley rats fed milk powder containing recombinant human lactoferrin (rhLF) derived from transgenic cloned cattle.

Science.gov (United States)

Zhou, Cui; Wang, Jian Wu; Huang, Kun Lun; He, XiaoYun; Chen, Xiu Ping; Sun, Hong; Yu, Tian; Che, Hui Lian

2011-10-01

Transgenic cloned animals expressing beneficial human nutritional traits offer a new strategy for large-scale production of some kinds of functional substances. In some cases, the required safety testing for genetically modified (GM) foods do not seem appropriate for human food safety, though regulations do not seem to provide alternatives. A 90-day rat feeding study is the core study for the safety assessment of GM foods. The test material in this 90-day study was prepared nonfat milk powder containing recombinant human lactoferrin (rhLF), which was expressed in transgenic cloned cattle. Groups of 10 male and female Sprague-Dawley rats were given a nutritionally balanced purified diet containing 7.5, 15, or 30% transgenic or conventional milk powder for 90 days. A commercial AIN93G diet was used as an additional control group. Clinical, biological, and pathological parameters were compared between groups. The only significant effect of treatment was higher mean ferritin and Fe(+) concentrations for both male and female rats fed the transgenic milk powder diets, as compared to rats fed nontransgenic milk diets or the commercial diet. The results of the present study are consistent with previous research, which indicates that milk powder containing rhLF derived from healthy transgenic cloned cattle is as safe as conventional milk powder.

Safety assessment of primary system components at the USNRC

Energy Technology Data Exchange (ETDEWEB)

Serpan, C Z; Chen, C Y; Taboada, A

1988-12-31

This document deals with the safety assessment in nuclear reactor components at the USNRC. The USNRC regulations and requirements concerning nuclear reactor design and operations are presented, together with guides and standards which describe how the actions should be implemented. The safety assessment relies on fracture analysis and Non Destructive Examination (NDE). (TEC).

Operational safety review programmes for nuclear power plants. Guidelines for assessment

International Nuclear Information System (INIS)

2002-01-01

The IAEA has been offering the Operational Safety Review Team (OSART) programme to provide advice and assistance to Member States in enhancing the operational safety of nuclear power plants (NPPs). Simultaneously, the IAEA has encouraged self-assessment and review by Member States of their own nuclear power plants to continuously improve nuclear safety. Currently, some utilities have been implementing safety review programmes to independently review their own plants. Corporate or national operational safety review programmes may be compliance or performance based. Successful utilities have found that both techniques are necessary to provide assurance that (i) as a minimum the NPP meets specific corporate and legal requirements and (ii) management at the NPP is encouraged to pursue continuous improvement principles. These programmes can bring nuclear safety benefits to the plants and utilities. The IAEA has conducted two pilot missions to assess the effectiveness of the operational review programme. Based on these missions and on the experience gained during OSART missions, this document has been developed to provide guidance on and broaden national/corporate safety review programmes in Member States, and to assist in maximizing their benefits. These guidelines are intended primarily for the IAEA team to conduct assessment of a national/corporate safety review programme. However, this report may also be used by a country or utility to establish its own national/corporate safety review programme. The guidelines may likewise be used for self-assessment or for establishing a baseline when benchmarking other safety review programmes. This report consists of four parts. Section 2 addresses the planning and preparation of an IAEA assessment mission and Sections 3 and 4 deal with specific guidelines for conducting the assessment mission itself

Seismic damage assessment of reinforced concrete containment structures

International Nuclear Information System (INIS)

Cho, HoHyun; Koh, Hyun-Moo; Hyun, Chang-Hun; Kim, Moon-Soo; Shin, Hyun Mock

2003-01-01

This paper presents a procedure for assessing seismic damage of concrete containment structures using the nonlinear time-history numerical analysis. For this purpose, two kinds of damage index are introduced at finite element and structural levels. Nonlinear finite element analysis for the containment structure applies PSC shell elements using a layered approach leading to damage indices at finite element and structural levels, which are then used to assess the seismic damage of the containment structure. As an example of such seismic damage assessment, seismic damages of the containment structure of Wolsong I nuclear power plant in Korea are evaluated against 30 artificial earthquakes generated with a wide range of PGA according to US NRC regulatory guide 1.60. Structural responses and corresponding damage index according to the level of PGA and nonlinearity are investigated. It is also shown that the containment structure behaves elastically for earthquakes corresponding to or lower than DBE. (author)

An update on safety and immunogenicity of vaccines containing emulsion-based adjuvants.

Science.gov (United States)

Fox, Christopher B; Haensler, Jean

2013-07-01

With the exception of alum, emulsion-based vaccine adjuvants have been administered to far more people than any other adjuvant, especially since the 2009 H1N1 influenza pandemic. The number of clinical safety and immunogenicity evaluations of vaccines containing emulsion adjuvants has correspondingly mushroomed. In this review, the authors introduce emulsion adjuvant composition and history before detailing the most recent findings from clinical and postmarketing data regarding the effects of emulsion adjuvants on vaccine immunogenicity and safety, with emphasis on the most widely distributed emulsion adjuvants, MF59® and AS03. The authors also present a summary of other emulsion adjuvants in clinical development and indicate promising avenues for future emulsion-based adjuvant development. Overall, emulsion adjuvants have demonstrated potent adjuvant activity across a number of disease indications along with acceptable safety profiles.

Environmental, safety, and health engineering

International Nuclear Information System (INIS)

Woodside, G.; Kocurek, D.

1997-01-01

A complete guide to environmental, safety, and health engineering, including an overview of EPA and OSHA regulations; principles of environmental engineering, including pollution prevention, waste and wastewater treatment and disposal, environmental statistics, air emissions and abatement engineering, and hazardous waste storage and containment; principles of safety engineering, including safety management, equipment safety, fire and life safety, process and system safety, confined space safety, and construction safety; and principles of industrial hygiene/occupational health engineering including chemical hazard assessment, personal protective equipment, industrial ventilation, ionizing and nonionizing radiation, noise, and ergonomics

The practice of pre-marketing safety assessment in drug development.

Science.gov (United States)

Chuang-Stein, Christy; Xia, H Amy

2013-01-01

The last 15 years have seen a substantial increase in efforts devoted to safety assessment by statisticians in the pharmaceutical industry. While some of these efforts were driven by regulations and public demand for safer products, much of the motivation came from the realization that there is a strong need for a systematic approach to safety planning, evaluation, and reporting at the program level throughout the drug development life cycle. An efficient process can help us identify safety signals early and afford us the opportunity to develop effective risk minimization plan early in the development cycle. This awareness has led many pharmaceutical sponsors to set up internal systems and structures to effectively conduct safety assessment at all levels (patient, study, and program). In addition to process, tools have emerged that are designed to enhance data review and pattern recognition. In this paper, we describe advancements in the practice of safety assessment during the premarketing phase of drug development. In particular, we share examples of safety assessment practice at our respective companies, some of which are based on recommendations from industry-initiated working groups on best practice in recent years.

Training courses on integrated safety assessment modelling for waste repositories

International Nuclear Information System (INIS)

Mallants, D.

2007-01-01

Near-surface or deep repositories of radioactive waste are being developed and evaluated all over the world. Also, existing repositories for low- and intermediate-level waste often need to be re-evaluated to extend their license or to obtain permission for final closure. The evaluation encompasses both a technical feasibility as well as a safety analysis. The long term safety is usually demonstrated by means of performance or safety assessment. For this purpose computer models are used that calculate the migration of radionuclides from the conditioned radioactive waste, through engineered barriers to the environment (groundwater, surface water, and biosphere). Integrated safety assessment modelling addresses all relevant radionuclide pathways from source to receptor (man), using in combination various computer codes in which the most relevant physical, chemical, mechanical, or even microbiological processes are mathematically described. SCK-CEN organizes training courses in Integrated safety assessment modelling that are intended for individuals who have either a controlling or supervising role within the national radwaste agencies or regulating authorities, or for technical experts that carry out the actual post-closure safety assessment for an existing or new repository. Courses are organised by the Department of Waste and Disposal

Aging characteristics of containment building and sensitivity on ultimate pressure capacity

International Nuclear Information System (INIS)

Seo, Jeong Moon; Choun, Young Sun; Choi, In Kil; Ha, Jae Joo

1998-03-01

For the reliable safety assessment of the containment building, structural and material conditions can be investigated in detail and pertinent assessment technologies have to be established. Also, an understanding on the aging-related degradations for the construction materials is required to predict long-term structural safety of the containment building. For the development of reliable aging prediction models, an extensive data base system related to aging properties of the containment building has to be prepared. The objectives of this research are to develop aging models representing long-term degradation of materials and a structural performance assessment program for containment building considering aging-related degradation. According to the results of sensitivity analysis, as the mechanical properties of the constituent materials degrade, the ultimate pressure capacity of containment building may decrease and severe damage may occur around the mid-level of the containment wall. (author). 28 refs., 11 tabs., 36 figs

Human Reliability in Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs

Interim process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Sellin, Patrick

2004-08-01

This report is a documentation of buffer processes identified as relevant to the long-term safety of a KBS-3 repository. The report is part of the interim reporting of the safety assessment SR-Can, see further the Interim main report. The final SR-Can reporting will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of this report is to document the scientific knowledge of the processes to a level required for an adequate treatment in the safety assessment. The documentation is thus from a scientific point of not exhaustive since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. The purpose is further to determine the handling of each process in the safety assessment and to demonstrate how uncertainties are taken care of, given the suggested handling. The process documentation in the SR 97 version of the Process report is a starting point for this SR-Can interim version. As further described in the Interim main report, the list of relevant processes has been reviewed and slightly extended by comparison to other databases. Furthermore, the backfill has been included as a system part of its own, rather than being described together with the buffer as in SR 97. Apart from giving an interim account of the documentation and handling of buffer processes in SR-Can, this report is meant to serve as a template for the forthcoming documentation of processes occurring in other parts of the repository system. A complete list of processes can be found in the Interim FEP report for the safety assessment SR-Can. All material presented in this document is preliminary in nature and will possibly be updated as the SR-Can project progresses

Interim process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrick (ed.)

2004-08-01

This report is a documentation of buffer processes identified as relevant to the long-term safety of a KBS-3 repository. The report is part of the interim reporting of the safety assessment SR-Can, see further the Interim main report. The final SR-Can reporting will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of this report is to document the scientific knowledge of the processes to a level required for an adequate treatment in the safety assessment. The documentation is thus from a scientific point of not exhaustive since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. The purpose is further to determine the handling of each process in the safety assessment and to demonstrate how uncertainties are taken care of, given the suggested handling. The process documentation in the SR 97 version of the Process report is a starting point for this SR-Can interim version. As further described in the Interim main report, the list of relevant processes has been reviewed and slightly extended by comparison to other databases. Furthermore, the backfill has been included as a system part of its own, rather than being described together with the buffer as in SR 97. Apart from giving an interim account of the documentation and handling of buffer processes in SR-Can, this report is meant to serve as a template for the forthcoming documentation of processes occurring in other parts of the repository system. A complete list of processes can be found in the Interim FEP report for the safety assessment SR-Can. All material presented in this document is preliminary in nature and will possibly be updated as the SR-Can project progresses.

Probabilistic safety assessment goals in Canada

International Nuclear Information System (INIS)

Snell, V.G.

1986-01-01

CANDU safety philosphy, both in design and in licensing, has always had a strong bias towards quantitative probabilistically-based goals derived from comparative safety. Formal probabilistic safety assessment began in Canada as a design tool. The influence of this carried over later on into the definition of the deterministic safety guidelines used in CANDU licensing. Design goals were further developed which extended the consequence/frequency spectrum of 'acceptable' events, from the two points defined by the deterministic single/dual failure analysis, to a line passing through lower and higher frequencies. Since these were design tools, a complete risk summation was not necessary, allowing a cutoff at low event frequencies while preserving the identification of the most significant safety-related events. These goals gave a logical framework for making decisions on implementing design changes proposed as a result of the Probabilistic Safety Analysis. Performing this analysis became a regulatory requirement, and the design goals remained the framework under which this was submitted. Recently, there have been initiatives to incorporate more detailed probabilistic safety goals into the regulatory process in Canada. These range from far-reaching safety optimization across society, to initiatives aimed at the nuclear industry only. The effectiveness of the latter is minor at very low and very high event frequencies; at medium frequencies, a justification against expenditures per life saved in other industries should be part of the goal setting

A regulatory view of containment integrity in the United Kingdom

Energy Technology Data Exchange (ETDEWEB)

Bradford, P M; Patchett, C M [Health and Safety Executive, Bootle (United Kingdom). Nuclear Installations Inspectorate

1994-12-31

This paper reviews the approach of HM Nuclear Installations Inspectorate (NII) to containment integrity in the United Kingdom (UK). NII is that part of the regulatory authority, the Health and Safety Executive (HSE), which administers the UK`s nuclear site licensing system. A major part of the licensing process lies in the assessment of licensees` submissions for new and existing plant. The purpose of this paper is to: briefly review our revised Safety Assessment Principles, describe our assessment and inspection activities on the primary containment building of the Sizewell B PWR which is progressing to full power operation in 1994 and, to indicate our views on the possible directions for future research into containment design and performance. (author). 5 refs.

Probabilistic Safety Assessment of Waste from PyroGreen Processes

International Nuclear Information System (INIS)

Ju, Hee Jae; Ham, In hye; Hwang, Il Soon

2016-01-01

The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

Probabilistic Safety Assessment of Waste from PyroGreen Processes

Energy Technology Data Exchange (ETDEWEB)

Ju, Hee Jae; Ham, In hye; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

2016-05-15

The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

Electronuclear's safety culture assessment and enhancement program

International Nuclear Information System (INIS)

Selvatici, E.; Diaz-Francisco, J.M.; Diniz de Souza, V.

2002-01-01

The present paper describes the Eletronuclear's safety culture assessment and enhancement program. The program was launched by the company's top management one year after the creation of Eletronuclear in 1997, from the merging of two companies with different organizational cultures, the design and engineering company Nuclen and the nuclear directorate of the Utility Furnas, Operator of the Angra1 NPP. The program consisted of an assessment performed internally in 1999 with the support and advice of the IAEA. This assessment, performed with the help of a survey, pooled about 80% of the company's employees. The overall result of the assessment was that a satisfactory level of safety culture existed; however, a number of points with a considerable margin for improvement were also identified. These points were mostly related with behavioural matters such as motivation, stress in the workplace, view of mistakes, handling of conflicts, and last but not least a view by a considerable number of employees that a conflict between safety and production might exist. An Action Plan was established by the company managers to tackle these weak points. This Plan was issued as company guideline by the company's Directorate. The subsequent step was to detail and implement the different actions of the Plan, which is the phase that we are at present. In the detailing of the Action Plan, special care was taken to sum up efforts, avoiding duplication of work or competition with already existing programs. In this process it was identified that the company had a considerable number of initiatives directly related to organizational and safety culture improvement, already operational. These initiatives have been integrated in the detailed Action Plan. A new assessment, for checking the effectiveness of the undertaken actions, is planned for 2003. (author)

The regulatory approach to the assessment of containment structures in the United Kingdom, with reference to the proposed Sizewell 'B' project

International Nuclear Information System (INIS)

Todd, I.W.

1988-01-01

The regulation of nuclear installations in the United Kingdom is carried out by means of a licensing system which is administered on behalf of the Health and Safety Executive (HSE) by HM Nuclear Installations Inspectorate. A major part of this process lies in the assessment of an installation's designs and safety case, which must be submitted to the Inspectorate before a licence may be granted. This paper describes the work carried out in assessing the design proposed for the containment structure of the United Kingdom's first PWR, Sizewell 'B'. The roles of the Inspectorate's Safety Assessment Principles, and the engineering judgement of the individual assessor, are discussed. The reasons behind the Inspectorate's decision to commission independent design studies of the structure are explained, and the results of these studies are presented in outline. Views are also expressed upon the validation of analytical results, quality assurance, and further work planned by the Inspectorate in this field. The paper concludes that it is the flexibility and discretion allowed by the above approach which represents the principal strength of the UK's system of regulatory assessment. (author)

Development of a quality assurance safety assessment database for near surface radioactive waste disposal

International Nuclear Information System (INIS)

Park, J. W.; Kim, C. L.; Park, J. B.; Lee, E. Y.; Lee, Y. M.; Kang, C. H.; Zhou, W.; Kozak, M. W.

2003-01-01

A quality assurance safety assessment database, called QUARK (QUality Assurance program for Radioactive waste management in Korea), has been developed to manage both analysis information and parameter database for safety assessment of Low- and Intermediate-Level radioactive Waste (LILW) disposal facility in Korea. QUARK is such a tool that serves QA purposes for managing safety assessment information properly and securely. In QUARK, the information is organized and linked to maximize the integrity of information and traceability. QUARK provides guidance to conduct safety assessment analysis, from scenario generation to result analysis, and provides a window to inspect and trace previous safety assessment analysis and parameter values. QUARK also provides default database for safety assessment staff who construct input data files using SAGE(Safety Assessment Groundwater Evaluation), a safety assessment computer code

Safety assessments for deep geological disposal of radioactive wastes

International Nuclear Information System (INIS)

Lyon, R.B.

1984-01-01

The objective of safety assessment for deep geological disposal of radioactive wastes is to evaluate how well the engineered barriers and geological setting inhibit radionuclide migration and prevent radiation dose to man. Safety assessment is influenced through interaction with the regulatory agencies, research groups, the public and the various levels of government. Under the auspices of the IAEA, a generic disposal system description has been developed to facilitate international exchange and comparison of data and results, and to enable development and comparison of performance for all components of the disposal system. It is generally accepted that a systems modelling approach is required and that safety assessment can be considered on two levels. At the systems level, all components of the system are taken into account to evaluate the risk to man. At the systems level, critical review and quality assurance on software provide the major validation techniques. Risk is a combination of dose estimate and probability of that dose. For analysis of the total system to be practical, the components are usually represented by simplified models. Recently, assessments have been taking uncertainties in the input data into account. At the detailed level, large-scale, complex computer programs model components of the system in sufficient detail that validation by comparison with field and laboratory measurements is possible. For example, three-dimensional fluid-flow, heat-transport and solute-transport computer programs have been used. Approaches to safety assessment are described, with illustrations from safety assessments performed in a number of countries. (author)

New safety performance indicators for safety assessment of radioactive waste disposal facilities. Cuban experience

International Nuclear Information System (INIS)

Peralta Vital, J.L.; Castillo, R.G.; Olivera, J.

2002-01-01

The paper shows the Cuban experience on implementing geological disposal of radioactive waste and the necessity for identifying new safety performance indicators for the safety assessment (SA) of radioactive waste disposal facilities. The selected indicator was the concentration of natural radioactive elements (U, Ra, Th, K) in the Cuban geologic environment. We have carried out a group of investigations, which have allowed characterising the concentration for the whole Country, creating a wide database where this indicator is associated with the lithology. The main lithologies in Cuba are: the sedimentary rocks (70 percent of national occurrence), which are present in the three regions (limestone and lutite), and finally the igneous and metamorphic rocks. The results show the concentrations ranges of the natural radionuclides associated fundamentally to the variation in the lithology and geographical area of the Country. In Cuba, the higher concentration (ppm) of Uranium and Radium are referenced to the Central region associated to Skarn, while for Thorium (ppm) and Potassium (%), in the East region the concentration peaks in Tuffs have been found. The concentrations ranges obtained are preliminary, they characterise the behaviour of this parameter for the Cuban geology, but they do not represent limits for safety assessment purposes yet. Also other factors should be taken into account as the assessment context, time scales and others assumptions before establishing the final concentration limits for the natural radionuclides as a radiological and nuclear safety performance indicator complementary to dose and risk for safety assessment for radiological and nuclear facilities. (author)

Remote-Reading Safety and Safeguards Surveillance System for 3013 Containers

International Nuclear Information System (INIS)

Lechelt, W. M.; Skorpik, J. R.; Silvers, K. L.; Szempruch, R. W.; Douglas, D. G.; Fein, K. O.

2002-01-01

At Hanford's Plutonium Finishing Plant (PFP), plutonium oxide is being loaded into stainless steel containers for long-term storage on the Hanford Site. These containers consist of two weld-sealed stainless steel cylinders nested one within the other. A third container holds the plutonium within the inner cylinder. This design meets the U.S. Department of Energy (DOE) storage standard, DOE-STD- 3013-2000, which anticipates a 50-year storage lifetime. The 3013 standard also requires a container surveillance program to continuously monitor pressure and to assure safeguards are adequate. However, the configuration of the container system makes using conventional measurement and monitoring methods difficult. To better meet the 3013 monitoring requirements, a team from Fluor Hanford (who manages the PFP), Pacific Northwest National Laboratory (PNNL), and Vista Engineering Technologies, LLC, developed a safer, cost-efficient, remote PFP 3013 container surveillance system. This new surveillance system is a combination of two successfully deployed technologies: (1) a magnetically coupled pressure gauge developed by Vista Engineering and (2) a radio frequency (RF) tagging device developed by PNNL. This system provides continuous, 100% monitoring of critical parameters with the containers in place, as well as inventory controls. The 3013 container surveillance system consists of three main elements: (1) an internal magnetic pressure sensor package, (2) an instrument pod (external electronics package), and (3) a data acquisition storage and display computer. The surveillance system described in this paper has many benefits for PFP and DOE in terms of cost savings and reduced personnel exposure. In addition, continuous safety monitoring (i.e., internal container pressure and temperature) of every container is responsible nuclear material stewardship and fully meets and exceeds DOE's Integrated Surveillance Program requirements

Safety assessment of consumption of glabrous canary seed (Phalaris canariensis L.) in rats.

Science.gov (United States)

Magnuson, B A; Patterson, C A; Hucl, P; Newkirk, R W; Ram, J I; Classen, H L

2014-01-01

Canary seed is a nutrient-rich cereal grain; however, it has not been used in human food in part due to concerns regarding safety of consumption. Glabrous or hairless canary seed has potential human food use as trichomes are absent. The objective of the oral feeding studies reported here was to assess the safety of yellow and brown glabrous canary seed cultivars as human cereal foods. The first study was a 90-day rat oral toxicity study, which compared the effects of diets containing 50% of either brown dehulled glabrous, brown hulled glabrous, or brown hulled pubescent (hairy) hulled canary seed to a diet containing 50% wheat. No significant adverse effects were observed. In a 28-day and a 90-day study rats were fed yellow or brown glabrous canary seed groats in the AIN-76 diet at concentrations levels of 2.5%, 5% and 10%. The NOAELs in 90-day study were 5.15 g/kg/d and 5.23 g/kg/d for yellow and brown canary seed groats. Consumption of canary seed was associated with reduced incidence and severity of liver lipidosis as compared to controls. The combined results of these studies clearly demonstrate the safety of consumption of glabrous canary seed, and support its use as a human cereal grain. Copyright © 2013 Elsevier Ltd. All rights reserved.

EFFICIENT QUANTITATIVE RISK ASSESSMENT OF JUMP PROCESSES: IMPLICATIONS FOR FOOD SAFETY

OpenAIRE

Nganje, William E.

1999-01-01

This paper develops a dynamic framework for efficient quantitative risk assessment from the simplest general risk, combining three parameters (contamination, exposure, and dose response) in a Kataoka safety-first model and a Poisson probability representing the uncertainty effect or jump processes associated with food safety. Analysis indicates that incorporating jump processes in food safety risk assessment provides more efficient cost/risk tradeoffs. Nevertheless, increased margin of safety...

Safety re-assessment of AECL test and research reactors

International Nuclear Information System (INIS)

Winfield, D.J.

1990-01-01

Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

Safety-analysis report for packaging (SARP) general-purpose heat-source module 750-Watt shipping container

International Nuclear Information System (INIS)

Whitney, M.A.; Burgan, C.E.; Blauvelt, R.K.; Zocher, R.W.; Bronisz, S.E.

1981-01-01

The SARP includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. Extensive tests and evaluations were performed to show that the container will function effectively with respect to all required standards and when subjected to normal transportation conditions and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion). In addition, a steady state temperature profile and radiation profile were measured using two heat sources that very closely resemble the GPHS. This gave an excellent representation of the GPHS temperature and radiation profile. A nuclear criticality safety analysis determined that all safety requirements are met

Value-impact assessment of safety-related modifications

International Nuclear Information System (INIS)

Knowles, W.M.C.; Dinnie, K.S.; Gordon, C.W.

1992-01-01