Climate and climate-related issues for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Naeslund, Jens-Ove (comp.)

2006-11-15

The purpose of this report is to document current scientific knowledge of the climate-related conditions and processes relevant to the long-term safety of a KBS-3 repository to a level required for an adequate treatment in the safety assessment SR-Can. The report also includes a concise background description of the climate system. The report includes three main chapters: A description of the climate system (Chapter 2); Identification and discussion of climate-related issues (Chapter 3); and, A description of the evolution of climate-related conditions for the safety assessment (Chapter 4). Chapter 2 includes an overview of present knowledge of the Earth climate system and the climate conditions that can be expected to occur in Sweden on a 100,000 year time perspective. Based on this, climate-related issues relevant for the long-term safety of a KBS-3 repository are identified. These are documented in Chapter 3 'Climate-related issues' to a level required for an adequate treatment in the safety assessment. Finally, in Chapter 4, 'Evolution of climate-related conditions for the safety assessment' an evolution for a 120,000 year period is presented, including discussions of identified climate-related issues of importance for repository safety. The documentation is from a scientific point of view not exhaustive, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of a safety assessment. As further described in the SR-Can Main Report and in the Features Events and Processes report, the content of the present report has been audited by comparison with FEP databases compiled in other assessment projects. This report follows as far as possible the template for documentation of processes regarded as internal to the repository system. However, the term processes is not used in this report, instead the term issue has been used. Each issue includes a set of processes together resulting in the

Model summary report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Vahlund, Fredrik

2006-10-15

This document is the model summary report for the safety assessment SR-Can. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SR-Can, a number of different computer codes are used. In order to better understand how these codes are related Assessment Model Flowcharts, AMFs, have been produced within the project. From these, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A large number of different computer codes are used in the assessment of which some are commercial while others are developed especially for the current assessment project. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: It must be demonstrated that the code is suitable for its purpose; It must be demonstrated that the code has been properly used; and, It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met.

Model summary report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Vahlund, Fredrik

2006-10-01

This document is the model summary report for the safety assessment SR-Can. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SR-Can, a number of different computer codes are used. In order to better understand how these codes are related Assessment Model Flowcharts, AMFs, have been produced within the project. From these, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A large number of different computer codes are used in the assessment of which some are commercial while others are developed especially for the current assessment project. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: It must be demonstrated that the code is suitable for its purpose; It must be demonstrated that the code has been properly used; and, It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

An Assessment of SKB's Performance Assessment Calculations in the Interim Main Report for the Safety Assessment SR-Can

International Nuclear Information System (INIS)

Maul, Philip; Robinson, Peter

2005-03-01

SKB have published their Interim Main Report of the safety assessment SR-Can, which is intended to establish the framework for what will be submitted in 2006 in support of a licence application for construction of the spent fuel encapsulation plant. This follows on from the SR-Can Planning Document published in 2003. The purpose of the Interim Report is stated to be to demonstrate the methodology that will be used for safety assessment. The present report evaluates the information provided in the Interim SR-Can Report that is relevant to the Performance Assessment (PA) calculations that SKB intend to undertake, using independent calculations to facilitate this process. SKB consider that the primary safety function is to isolate completely the fuel within the canisters over the entire assessment period. Should a canister be damaged, the secondary safety function is to ensure that any release is retarded and dispersed sufficiently to ensure that concentrations levels in the accessible environment cannot cause unacceptable consequences. In this report PA calculations are considered to include both a high-level representation of the evolution of the system (relevant to the primary safety function), and any subsequent radionuclide transport (relevant to the secondary safety function). The main conclusions drawn are: 1. The effects of climate evolution on engineered barriers have not been analysed in detail in the Interim Report, and this limits the usefulness of the preliminary calculations that have been undertaken. 2. A key aspect of SKB's approach is the use of an integrated near-field evolution model. The information provided on this model demonstrates its capability efficiently to reproduce calculations from individual process models, but insufficient information is given at the present time to justify statements about interactions between processes. In particular it is assumed that relatively short term thermal and resaturation processes do not affect the

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

Data report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Vahlund, Fredrik; Andersson, Johan; Loefgren, Martin

2006-11-01

This report is the data report derived within the project SR-Can. The purpose of the data report is to present input data, with uncertainty estimates, for the SR-Can assessment calculations. Data presented in the report have been derived using standardised procedures following a methodology which is presented in the initial part of the report. In this part, a template is presented that has been used when assessing input data in supporting documents as illustrated in subsequent chapters of the data report. By using the template, decisions by the SR-Can team are separated from expert input. This increases the traceability of assessment decisions. The data report supplies assessment data for all parts of the repository system, the fuel, the canister, the buffer and backfill and the geosphere. For the geosphere, many of the data are based on information obtained during the site investigation programme

Data report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Vahlund, Fredrik [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Andersson, Johan [JA Streamflow AB, Aelvsjoe (Sweden); Loefgren, Martin [Kemakta Konsult AB, Stockholm (Sweden)

2006-11-15

This report is the data report derived within the project SR-Can. The purpose of the data report is to present input data, with uncertainty estimates, for the SR-Can assessment calculations. Data presented in the report have been derived using standardised procedures following a methodology which is presented in the initial part of the report. In this part, a template is presented that has been used when assessing input data in supporting documents as illustrated in subsequent chapters of the data report. By using the template, decisions by the SR-Can team are separated from expert input. This increases the traceability of assessment decisions. The data report supplies assessment data for all parts of the repository system, the fuel, the canister, the buffer and backfill and the geosphere. For the geosphere, many of the data are based on information obtained during the site investigation programme.

International Expert Review of Sr-Can: Safety Assessment Methodology - External review contribution in support of SSI's and SKI's review of SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sagar, Budhi (Center for Nuclear Waste Regulatory Analyses, Southwest Research Inst., San Antonio, TX (US)); Egan, Michael (Quintessa Limited, Henley-on-Thames (GB)); Roehlig, Klaus-Juergen (Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (DE)); Chapman, Neil (Independent Consultant (XX)); Wilmot, Roger (Galson Sciences Limited, Oakham (GB))

2008-03-15

In 2006, SKB published a safety assessment (SR-Can) as part of its work to support a licence application for the construction of a final repository for spent nuclear fuel. The purposes of the SR-Can project were stated in the main project report to be: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's research and development (R and D) programme, to further site investigations and to future safety assessments. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. To help inform their review of SKB's proposed approach to development of the longterm safety case, the authorities appointed three international expert review teams to carry out a review of SKB's SR-Can safety assessment report. Comments from one of these teams - the Safety Assessment Methodology (SAM) review team - are presented in this document. The SAM review team's scope of work included an examination of SKB's documentation of the assessment ('Long-term safety for KBS-3 Repositories at Forsmark and Laxemar - a first evaluation' and several supporting reports) and hearings with SKB staff and contractors, held in March 2007. As directed by SKI and SSI, the SAM review team focused on methodological aspects and sought to determine whether SKB's proposed safety assessment methodology is likely to be suitable for use in the future SR-Site and to assess its consistency with the Swedish regulatory framework. No specific evaluation of long-term safety or site acceptability was undertaken by any of the review teams. SKI and SSI's Terms of Reference for the SAM

SKI's and SSI's review of SKB's safety report SR-Can

International Nuclear Information System (INIS)

Dverstorp, Bjoern; Stroemberg, Bo

2008-03-01

This report summarises SKI's and SSI's joint review of the Swedish Nuclear Fuel and Waste Management Co's (SKB) safety report SR-Can (SKB TR-06-09). SR-Can is the first assessment of post-closure safety for a KBS-3 spent nuclear fuel repository at the candidate sites Forsmark and Laxemar, respectively. The analysis builds on data from the initial stage of SKB's surface-based site investigations and on data from full-scale manufacturing and testing of buffer and copper canisters. SR-Can can be regarded as a preliminary version of the safety report that will be required in connection with SKB's planned licence application for a final repository in late 2009. The main purpose of the authorities' review is to provide feedback to SKB on their safety reporting as part of the pre-licensing consultation process. However, SR-Can is not part of the formal licensing process. In support of the authorities' review three international peer review teams were set up to make independent reviews of SR-Can from three perspectives, namely integration of site data, representation of the engineered barriers and safety assessment methodology, respectively. Further, several external experts and consultants have been engaged to review detailed technical and scientific issues in SR-Can. The municipalities of Oesthammar and Oskarshamn where SKB is conducting site investigations, as well NGOs involved in SKB's programme, have been invited to provide their views on SR-Can as input to the authorities' review. Finally, the authorities themselves, and with the help of consultants, have used independent models to reproduce part of SKB's calculations and to make complementary calculations. All supporting review documents are published in SKI's and SSI's report series. The main findings of the review are: -SKB's safety assessment methodology is overall in accordance with applicable regulations, but part of the methodology needs to be further developed for the licence application. -SKB's quality

Interim data report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Vahlund, Fredrik; Andersson, Johan

2004-08-01

This document is the interim data report in the project SR-Can. The purpose of the data report is to present input data, with uncertainty estimates, for the SR-Can assessment calculations. Besides input data, the report also describes the standardised procedures used when deriving the input data and the corresponding uncertainty estimates. However, in the present interim version of the report (written in the initial stage of the project when site characterisation has yet not been completed) the standardised procedures have not been possible to apply for most of the data and, in order to present a compilation of the data used in the assessment, much of the input data is presented without following the standardised procedures. This will however be changed for the final version of the SR-Can data report, in order to show the methodology that will be used in the final version one example of how input data will be presented is included (migration data for buffer) . The recommended input data for the assessment calculations are, for the interim version, mainly based on SR 97 Beberg data, these are merely presented without any background or uncertainty discussion (this is presented in the SR 97 data report)

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

SSI's independent consequence calculations in support of the regulatory review of the SR-Can safety assessment

International Nuclear Information System (INIS)

Shulan Xu; Dverstorp, Bjoern; Woerman, Anders; Marklund, Lars; Klos, Richard; Shaw, George

2008-03-01

With the publication of the SR-Can report at the end of 2006, Swedish Nuclear Fuel and Waste Management Co (SKB) have presented a complete assessment of long-term safety for a KBS-3 repository. The SR-Can project demonstrates progress in SKB's capabilities in respect of the methodology for assessment of long-term safety in support of a licence application for a final repository. According to SKB's plans, applications to construct a geological repository will be submitted in 2009, supported by post-closure safety assessments. Project CLIMB (Catchment LInked Models of radiological effects in the Biosphere) was instituted in 2004 to provide SSI with an independent modelling capability when reviewing SKB's assessments. Modelling in CLIMB covers all aspects of performance assessment (PA) from nearfield releases to radiological consequences in the surface environment. This review of SR-Can provides the first opportunity to apply the models and to compare the CLIMB approach with developments at SKB. The aim of the independent calculations is to investigate key aspects of the PA models and so to better understand the assessment methodology used by SKB. Independent modelling allows critical review issues to be addressed by the application of alternative models and assumptions. Three reviews are undertaken here: - Reproduction of selected cases from SR-Can in order to demonstrate an adequate understanding of the PA model from details given in the SR-Can documentation. - Alternative conceptualisation of radionuclide transport and accumulation in the surface system. Two modelling approaches have been used: GEMA (the Generic Ecosystem Modelling Approach) is a traditional compartmental model similar to that used by SKB in SR-Can but with additional functionality and flexibility. The second approach takes continuous transport models to investigate contaminant migration through the Quaternary deposits into the surface drainage system. - The final strand of the CLIMB investigation

Interim initial state report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Pers, Karin (ed.) [Kemakta Konsult AB, Stockholm (Sweden)

2004-07-01

A thorough description of the initial state of the engineered parts of the repository system is one of the main bases for the SR-Can safety assessment. The initial state refers to the state at the time of deposition for the spent fuel and the engineered barriers and the natural, undisturbed state at the time of beginning of excavation for the repository for the geosphere and the biosphere. The repository system is based on the KBS-3 method, where 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. For the purpose of the safety assessment the engineered portion of the repository system has been divided into a number of consecutive barriers or sub-systems. The importance of a particular feature for safety has influenced the resolution into components. In principle, components close to the source term and those that play an important role for safety are treated in more detail than more peripheral components. For the option with 40 years of reactor operation, the quantity of BWR fuel is estimated at 7200 tonnes and the quantity of PWR fuel at 2300 tonnes. The fuel burn-up may vary from 15 MWd/kgU up to 60 MWd/kg. Geometric aspects of the fuel cladding tubes of importance in the safety assessment are, as a rule, handled sufficiently pessimistically in analyses of radionuclide transport that differences between different fuel types are irrelevant. The relative differences in radionuclide inventory with respect to burn-up are small. Deviations in inventory and deviating or damaged fuel are not considered in the SR-Can interim reporting but will be handled in the final reporting of SR-Can. The canister consists of an inner container, the insert of cast iron and an outer shell of copper. The cast iron insert provides mechanical stability and the copper shell protects against corrosion in the repository environment. The copper shell is 5 cm thick and

Interim initial state report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Pers, Karin

2004-07-01

A thorough description of the initial state of the engineered parts of the repository system is one of the main bases for the SR-Can safety assessment. The initial state refers to the state at the time of deposition for the spent fuel and the engineered barriers and the natural, undisturbed state at the time of beginning of excavation for the repository for the geosphere and the biosphere. The repository system is based on the KBS-3 method, where 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. For the purpose of the safety assessment the engineered portion of the repository system has been divided into a number of consecutive barriers or sub-systems. The importance of a particular feature for safety has influenced the resolution into components. In principle, components close to the source term and those that play an important role for safety are treated in more detail than more peripheral components. For the option with 40 years of reactor operation, the quantity of BWR fuel is estimated at 7200 tonnes and the quantity of PWR fuel at 2300 tonnes. The fuel burn-up may vary from 15 MWd/kgU up to 60 MWd/kg. Geometric aspects of the fuel cladding tubes of importance in the safety assessment are, as a rule, handled sufficiently pessimistically in analyses of radionuclide transport that differences between different fuel types are irrelevant. The relative differences in radionuclide inventory with respect to burn-up are small. Deviations in inventory and deviating or damaged fuel are not considered in the SR-Can interim reporting but will be handled in the final reporting of SR-Can. The canister consists of an inner container, the insert of cast iron and an outer shell of copper. The cast iron insert provides mechanical stability and the copper shell protects against corrosion in the repository environment. The copper shell is 5 cm thick and

Review of SKB's Safety Assessment SR-Can: Contributions in Support of SKI's and SSI's Review by External Consultants

International Nuclear Information System (INIS)

2008-03-01

The Swedish Nuclear Fuel and Waste Management Co (SKB) plans to submit a license application for the construction of a repository for spent nuclear fuel in Sweden 2010. In support of this application SKB will present a safety report, SR-Site, on the repository's long-term safety and radiological consequences. As a preparation for SR-Site, SKB published the preliminary safety assessment SR-Can in November 2006. The purposes were to document a first evaluation of long-term safety for the two candidate sites at Forsmark and Laxemar and to provide feedback to SKB's future programme of work. An important objective of the authorities' review of SR-Can is to provide guidance to SKB on the complete safety reporting for the license application. The authorities have engaged external experts for independent modelling, analysis and review, with the aim to provide a range of expert opinions related to the sufficiency and appropriateness of various aspects of SR-Can. The conclusions and judgments in this report are those of the authors and may not necessarily coincide with those of SKI and SSI. The authorities own review will be published separately (SKI Report 2008:23, SSI Report 2008:04 E). This report compiles contributions from several specific research projects. The separate reviews cover topics regarding the engineered barrier system, the quality assurance, the climate evolution and its effects, and the ecosystems and environmental impacts. All contributions are in English apart from the review concerning ecosystems and environmental impacts, which is presented in Swedish

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

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/

SSI's independent consequence calculations in support of the regulatory review of the SR-Can safety assessment

Energy Technology Data Exchange (ETDEWEB)

Shulan Xu; Dverstorp, Bjoern (Swedish Radiation Protection Authority, Stockholm (Sweden)); Woerman, Anders; Marklund, Lars (Royal Institute of Technology (KTH), Stockholm (SE)); Klos, Richard (Aleksandria Sciences, Sheffield (GB)); Shaw, George (Univ. of Nottingham (GB))

2008-03-15

With the publication of the SR-Can report at the end of 2006, Swedish Nuclear Fuel and Waste Management Co (SKB) have presented a complete assessment of long-term safety for a KBS-3 repository. The SR-Can project demonstrates progress in SKB's capabilities in respect of the methodology for assessment of long-term safety in support of a licence application for a final repository. According to SKB's plans, applications to construct a geological repository will be submitted in 2009, supported by post-closure safety assessments. Project CLIMB (Catchment LInked Models of radiological effects in the Biosphere) was instituted in 2004 to provide SSI with an independent modelling capability when reviewing SKB's assessments. Modelling in CLIMB covers all aspects of performance assessment (PA) from nearfield releases to radiological consequences in the surface environment. This review of SR-Can provides the first opportunity to apply the models and to compare the CLIMB approach with developments at SKB. The aim of the independent calculations is to investigate key aspects of the PA models and so to better understand the assessment methodology used by SKB. Independent modelling allows critical review issues to be addressed by the application of alternative models and assumptions. Three reviews are undertaken here: - Reproduction of selected cases from SR-Can in order to demonstrate an adequate understanding of the PA model from details given in the SR-Can documentation. - Alternative conceptualisation of radionuclide transport and accumulation in the surface system. Two modelling approaches have been used: GEMA (the Generic Ecosystem Modelling Approach) is a traditional compartmental model similar to that used by SKB in SR-Can but with additional functionality and flexibility. The second approach takes continuous transport models to investigate contaminant migration through the Quaternary deposits into the surface drainage system. - The final strand of the CLIMB

Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

Energy Technology Data Exchange (ETDEWEB)

2009-11-15

The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

International Nuclear Information System (INIS)

2009-11-01

The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

Radionuclide transport report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This document compiles radionuclide transport calculations of a KBS-3 repository for the safety assessment SR-Site. The SR-Site assessment supports the licence application for a final repository at Forsmark, Sweden

Initial state report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Pers, Karin (ed.) [Kemakta Konsult AB, Stockholm (Sweden)

2006-10-15

A comprehensive description of the initial state of the engineered parts of the repository system is one of the main bases for the safety assessment. There is no obvious definition of the time of the initial state. For the engineered part of their repository system, the time of deposition is a natural starting point and the initial state in SR-Can is, therefore, defined as the state at the time of deposition for the engineered barrier system. The initial state of the engineered parts of the repository system is largely obtained from the design specifications of the repository, including allowed tolerances or allowance for deviations. Also the manufacturing, excavation and control methods have to be described in order to adequately discuss and handle hypothetical initial states outside the allowed limits in the design specifications. It should also be noted that many parts of the repository system are as yet not finally designed, there can be many changes in the future. The design and technical solutions presented here are representative of the current stage of development. The repository system is based on the KBS-3 method, in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at 400-700 m depth in saturated granitic rock. The facility design comprises rock caverns, tunnels, deposition positions etc. Deposition tunnels are linked by tunnels for transport and communication and shafts for ventilation. One ramp and five shafts connect the surface facility to the underground repository. The ramp is used for heavy and bulky transports and the shafts are for utility systems and for transport of excavated rock, backfill and staff. For the purposes of the safety assessment, the engineered parts of the repository system have been sub-divided into a number of components or sub-systems. These are: The fuel, (also including cavities in the canister since strong interactions between the two occur if the

Initial state report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Pers, Karin

2006-10-01

A comprehensive description of the initial state of the engineered parts of the repository system is one of the main bases for the safety assessment. There is no obvious definition of the time of the initial state. For the engineered part of their repository system, the time of deposition is a natural starting point and the initial state in SR-Can is, therefore, defined as the state at the time of deposition for the engineered barrier system. The initial state of the engineered parts of the repository system is largely obtained from the design specifications of the repository, including allowed tolerances or allowance for deviations. Also the manufacturing, excavation and control methods have to be described in order to adequately discuss and handle hypothetical initial states outside the allowed limits in the design specifications. It should also be noted that many parts of the repository system are as yet not finally designed, there can be many changes in the future. The design and technical solutions presented here are representative of the current stage of development. The repository system is based on the KBS-3 method, in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at 400-700 m depth in saturated granitic rock. The facility design comprises rock caverns, tunnels, deposition positions etc. Deposition tunnels are linked by tunnels for transport and communication and shafts for ventilation. One ramp and five shafts connect the surface facility to the underground repository. The ramp is used for heavy and bulky transports and the shafts are for utility systems and for transport of excavated rock, backfill and staff. For the purposes of the safety assessment, the engineered parts of the repository system have been sub-divided into a number of components or sub-systems. These are: The fuel, (also including cavities in the canister since strong interactions between the two occur if the

SKI's and SSI's review of SKB's safety report SR-Can

Energy Technology Data Exchange (ETDEWEB)

Dverstorp, Bjoern; Stroemberg, Bo (and others)

2008-03-15

This report summarises SKI's and SSI's joint review of the Swedish Nuclear Fuel and Waste Management Co's (SKB) safety report SR-Can (SKB TR-06-09). SR-Can is the first assessment of post-closure safety for a KBS-3 spent nuclear fuel repository at the candidate sites Forsmark and Laxemar, respectively. The analysis builds on data from the initial stage of SKB's surface-based site investigations and on data from full-scale manufacturing and testing of buffer and copper canisters. SR-Can can be regarded as a preliminary version of the safety report that will be required in connection with SKB's planned licence application for a final repository in late 2009. The main purpose of the authorities' review is to provide feedback to SKB on their safety reporting as part of the pre-licensing consultation process. However, SR-Can is not part of the formal licensing process. In support of the authorities' review three international peer review teams were set up to make independent reviews of SR-Can from three perspectives, namely integration of site data, representation of the engineered barriers and safety assessment methodology, respectively. Further, several external experts and consultants have been engaged to review detailed technical and scientific issues in SR-Can. The municipalities of Oesthammar and Oskarshamn where SKB is conducting site investigations, as well NGOs involved in SKB's programme, have been invited to provide their views on SR-Can as input to the authorities' review. Finally, the authorities themselves, and with the help of consultants, have used independent models to reproduce part of SKB's calculations and to make complementary calculations. All supporting review documents are published in SKI's and SSI's report series. The main findings of the review are: -SKB's safety assessment methodology is overall in accordance with applicable regulations, but part of the methodology needs to be

Fuel and canister process report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Werme, Lars; Lilja, Christina

2010-12-01

This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It 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/

Fuel and canister process report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Werme, Lars; Lilja, Christina (eds.)

2010-12-15

This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It 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/

Climate and climate-related issues for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

The purpose of this report is to document current scientific knowledge on climate and climate-related conditions, relevant to the long-term safety of a KBS-3 repository, to a level required for an adequate treatment in the safety assessment SR-Site. The report also presents a number of dedicated studies on climate and selected climate-related processes of relevance for the assessment of long term repository safety. Based on this information, the report presents a number of possible future climate developments for Forsmark, the site selected for building a repository for spent nuclear fuel in Sweden (Figure 1-1). The presented climate developments are used as basis for the selection and analysis of SR-Site safety assessment scenarios in the SR-Site main report /SKB 2011/. The present report is based on research conducted and published by SKB as well as on research reported in the general scientific literature

Climate and climate-related issues for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

The purpose of this report is to document current scientific knowledge on climate and climate-related conditions, relevant to the long-term safety of a KBS-3 repository, to a level required for an adequate treatment in the safety assessment SR-Site. The report also presents a number of dedicated studies on climate and selected climate-related processes of relevance for the assessment of long term repository safety. Based on this information, the report presents a number of possible future climate developments for Forsmark, the site selected for building a repository for spent nuclear fuel in Sweden (Figure 1-1). The presented climate developments are used as basis for the selection and analysis of SR-Site safety assessment scenarios in the SR-Site main report /SKB 2011/. The present report is based on research conducted and published by SKB as well as on research reported in the general scientific literature

Review of SKB's Safety Assessment SR-Can: Contributions in Support of SKI's and SSI's Review by External Consultants

Energy Technology Data Exchange (ETDEWEB)

2008-03-15

The Swedish Nuclear Fuel and Waste Management Co (SKB) plans to submit a license application for the construction of a repository for spent nuclear fuel in Sweden 2010. In support of this application SKB will present a safety report, SR-Site, on the repository's long-term safety and radiological consequences. As a preparation for SR-Site, SKB published the preliminary safety assessment SR-Can in November 2006. The purposes were to document a first evaluation of long-term safety for the two candidate sites at Forsmark and Laxemar and to provide feedback to SKB's future programme of work. An important objective of the authorities' review of SR-Can is to provide guidance to SKB on the complete safety reporting for the license application. The authorities have engaged external experts for independent modelling, analysis and review, with the aim to provide a range of expert opinions related to the sufficiency and appropriateness of various aspects of SR-Can. The conclusions and judgments in this report are those of the authors and may not necessarily coincide with those of SKI and SSI. The authorities own review will be published separately (SKI Report 2008:23, SSI Report 2008:04 E). This report compiles contributions from several specific research projects. The separate reviews cover topics regarding the engineered barrier system, the quality assurance, the climate evolution and its effects, and the ecosystems and environmental impacts. All contributions are in English apart from the review concerning ecosystems and environmental impacts, which is presented in Swedish

SKI's and SSI's joint review of the Swedish Nuclear Fuel and Waste Management Co's (SKB) safety report SR-Can; SKIs och SSIs gemensamma granskning av SKBs saekerhetsrapport SR-Can

Energy Technology Data Exchange (ETDEWEB)

Dverstorp, Bjoern; Stroemberg, Bo (and others)

2008-03-15

This report summarizes SKI's and SSI's joint review of the Swedish Nuclear Fuel and Waste Management Co's (SKB) safety report SR-Can (SKB TR-06-09). SR-Can is the first assessment of post-closure safety for a KBS-3 spent nuclear fuel repository at the candidate sites Forsmark and Laxemar, respectively. The analysis builds on data from the initial stage of SKB's surface-based site investigations and on data from full-scale manufacturing and testing of buffer and copper canisters. SR-Can can be regarded as a preliminary version of the safety report that will be required in connection with SKB's planned license application for a final repository in late 2009. The main purpose of the authorities' review is to provide feedback to SKB on their safety reporting as part of the pre-licensing consultation process. However, SR-Can is not part of the formal licensing process. In support of the authorities' review three international peer review teams were set up to make independent reviews of SR-Can from three perspectives, namely integration of site data, representation of the engineered barriers and safety assessment methodology, respectively. Further, several external experts and consultants have been engaged to review detailed technical and scientific issues in SR-Can. The municipalities of Oesthammar and Oskarshamn where SKB is conducting site investigations, as well NGOs involved in SKB's programme, have been invited to provide their views on SR-Can as input to the authorities' review. Finally, the authorities themselves, and with the help of consultants, have used independent models to reproduce part of SKB's calculations and to make complementary calculations. All supporting review documents are published in SKI's and SSI's report series. The main findings of the review are: SKB's safety assessment methodology is overall in accordance with applicable regulations, but part of the methodology needs to be

Review of SKB's interim report of SR-Can: SKI's and SSI's evaluation of SKB's up-dated methodology for safety assessment

International Nuclear Information System (INIS)

Dverstorp, Bjoern; Moberg, Leif; Wiebert, Anders; Xu Shulan; Stroemberg, Bo; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Sundstroem, Benny; Toverud, Oeivind

2005-07-01

This report presents the findings of a review of the Swedish Nuclear Fuel and Waste Management Co.'s (SKB) interim report of the safety assessment SR-Can (SKB TR 04-11), conducted by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). SKB's interim report describes and exemplifies the safety assessment methodology that SKB plans to use in the oncoming licence applications for an encapsulation plant and a final repository for spent nuclear fuel. The authorities' review takes into account the findings of an international peer review of SKB's interim report. The authorities conclude that SKB has improved its safety assessment methodology in several aspects compared to earlier safety reports. Among other things the authorities commend SKB for giving a comprehensive account of relevant regulations and guidance, and for the systematic approach to identification and documentation of features, events and processes that need to be considered in the safety assessment. However, the authorities also conclude that important parts of SKB's method need to be further developed before they are mature enough to be used as a basis for a license application. The authorities' overall assessment is summarised in chapter 8 of this report

SR-CAN - a safety assessment of a repository of spent nuclear fuel: canister performance and effects on the biosphere

International Nuclear Information System (INIS)

Kautsky, U.; Kumblad, L.

2004-01-01

During the next few years the Swedish Nuclear Fuel and Waste Management Co. (SKB) performs site investigations at two sites in Sweden for a future repository of spent nuclear fuel. Parallel an encapsulation plant is planned to encapsulate the spent fuel in copper canisters according to the KBS-3 method. The purpose of the SR-CAN safety assessment is to show the performance of the canister isolations at different sites for a repository at 500 meters depth in crystalline rock. Moreover, SR-CAN provides an example how the site specific safety assessment of a deep repository will be made in year 2006-2008. To be able to calculate dose and risk for humans and the environment, new assessment methods were developed for the biosphere. These methods were based on a system ecological approach and used knowledge from landscape ecology to provide an integrated approach with hydrology and geology considering the discharges in a watershed and calculating consequences in terrestrial and aquatic (freshwater and marine) ecosystems. A range of methods and tools were developed in GIS and Matlab/Simulink to be able to model and understand the important processes in the landscape today and during the next few thousands of years. In this paper, an overview of the program and the novel methods are presented, as well as some examples from performance calculations from a watershed in the Forsmark area considering effects on humans and ecosystems. (author)

Data report for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

Data report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

SR-can: preliminary feedback to canister fabrication, repository design and future R and D

International Nuclear Information System (INIS)

Hedin, A.; Sellin, P.

2007-01-01

This paper discusses preliminary feedback from SKB's on-going safety assessment SR-Can, to he finalized in 2006. The assessment, which is not part of a formal licence application, is an important step towards the SR-Site assessment to be delivered in 2008 and which will support a licence application for a Swedish deep repository for spent nuclear fuel. The SR-Can assessment will use data from the initial stage of the on-going site investigations at the two candidate sites at Forsmark and Oskarshamn. Review comments on SR-Can from Swedish authorities are expected in the summer of 2007 and these will be taken into account when preparing the SR-Site assessment. An Interim version of the SR-Can report was produced in September 2004 and has been reviewed by the Swedish authorities supported by an international review team. The assessment concerns a KBS 3 repository for which the key safety related features can be summarised in the primary safety function isolation and the secondary function retardation. The isolation function is more prominent in the KBS 3 method compared to many other repository concepts. (authors)

Model summary report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Vahlund, Fredrik; Zetterstroem Evins, Lena; Lindgren, Maria

2010-12-01

This document is the model summary report for the safety assessment SR-Site. In the report, the quality assurance (QA) measures conducted for assessment codes are presented together with the chosen QA methodology. In the safety assessment project SR-Site, a large number of numerical models are used to analyse the system and to show compliance. In order to better understand how the different models interact and how information are transferred between the different models Assessment Model Flowcharts, AMFs, are used. From these, different modelling tasks can be identify and the computer codes used. As a large number of computer codes are used in the assessment the complexity of these differs to a large extent, some of the codes are commercial while others are developed especially for the assessment at hand. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined for all codes: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. - It must be described how data are transferred between the different computational tasks. Although the requirements are identical for all codes in the assessment, the measures used to show that the requirements are fulfilled will be different for different types of codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented together with a discussion on how the requirements are met

Model summary report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Vahlund, Fredrik; Zetterstroem Evins, Lena (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Lindgren, Maria (Kemakta Konsult AB, Stockholm (Sweden))

2010-12-15

This document is the model summary report for the safety assessment SR-Site. In the report, the quality assurance (QA) measures conducted for assessment codes are presented together with the chosen QA methodology. In the safety assessment project SR-Site, a large number of numerical models are used to analyse the system and to show compliance. In order to better understand how the different models interact and how information are transferred between the different models Assessment Model Flowcharts, AMFs, are used. From these, different modelling tasks can be identify and the computer codes used. As a large number of computer codes are used in the assessment the complexity of these differs to a large extent, some of the codes are commercial while others are developed especially for the assessment at hand. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined for all codes: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. - It must be described how data are transferred between the different computational tasks. Although the requirements are identical for all codes in the assessment, the measures used to show that the requirements are fulfilled will be different for different types of codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented together with a discussion on how the requirements are met

Long-term safety for KBS-3 repositories at Forsmark and Laxemar - a first evaluation. Main Report of the SR-Can project

International Nuclear Information System (INIS)

Hedin, Allan

2006-10-01

This document is the main report from the safety assessment project SR-Can. The SR-Can project is a preparatory stage for the SR-Site assessment, the report that will be used in support of SKB's application for a final repository. The purposes of the safety assessment SR-Can are the following: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's RandD programme, to further site investigations and to future safety assessment projects. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. 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 and Laxemar sites, presently being investigated by SKB as candidates for a KBS-3 repository are used in the assessment. An important aim of this report is to demonstrate the proper handling of requirements placed on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Institute are reproduced in an Appendix where references are given to sections in the main text where the handling of the different requirements is discussed. 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. The risk limit corresponds to an

Long-term safety for KBS-3 repositories at Forsmark and Laxemar - a first evaluation. Main Report of the SR-Can project

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan (ed.)

2006-10-15

This document is the main report from the safety assessment project SR-Can. The SR-Can project is a preparatory stage for the SR-Site assessment, the report that will be used in support of SKB's application for a final repository. The purposes of the safety assessment SR-Can are the following: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's RandD programme, to further site investigations and to future safety assessment projects. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. 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 and Laxemar sites, presently being investigated by SKB as candidates for a KBS-3 repository are used in the assessment. An important aim of this report is to demonstrate the proper handling of requirements placed on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Institute are reproduced in an Appendix where references are given to sections in the main text where the handling of the different requirements is discussed. 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

Buffer, backfill and closure process report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Sellin, Patrik

2010-11-01

This report gives an account of how processes in buffer, deposition tunnel backfill and the closure important for the long-term evolution of a KBS-3 repository for spent nuclear fuel, will be documented in the safety assessment SR-Site

Buffer, backfill and closure process report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrik (ed.)

2010-11-15

This report gives an account of how processes in buffer, deposition tunnel backfill and the closure important for the long-term evolution of a KBS-3 repository for spent nuclear fuel, will be documented in the safety assessment SR-Site

Review of SR 97 performance assessment

International Nuclear Information System (INIS)

Glynn, P.D.

2000-01-01

This review has identified many technical problems in the SR 97 performance assessment. The general impression of this reviewer is that SKB has been disingenuous in its performance assessment effort. It has not cited important differences of opinion with its own views. Furthermore, there are many inconsistencies in the SR 97 report that all together leave the impression that there are many more uncertainties in the SR 97 performance assessment than SKB would perhaps care to admit. Additionally, despite SKB's statements to the contrary, many of the analyses conducted for the SR 97 performance assessment can be clearly shown not to have been based on 'conservative' assumptions. Finally, SKB has made little effort to consider possible coupling effects between their different scenarios in SR 97. This is a serious flaw in the SR 97 performance assessment. The comments in this review should not be taken to imply that the KBS-3 nuclear waste disposal method will not be able to meet the safety and radiation protection requirements which SKI and SSI have specified in recent years. Instead, my conclusion is simply that the SR 97 performance assessment of the KBS-3 method would have been more believable had it been based on a forthright and comprehensive discussion of facts, uncertainties and opinions, and on a more conservative choice of assumptions. As it stands, the SR 97 performance assessment is not very credible

Handling of future human actions in the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

authorities in their review of SR-Can /Dverstorp and Stroemberg 2008/ maintain that the state, rather than SKB, is expected to be responsible for the supervision and monitoring of the repository after sealing. Man is dependent on, and influences, the environment in which he lives. After the repository has been closed, future generations should be able to utilise the repository site according to their needs without jeopardising their health. In the case of a final repository of the KBS-3 type, there are, however, inevitably examples of activities that, if carried out carelessly or without knowledge of the repository, could result in exposure to radiotoxic elements from the spent fuel. Therefore, there is an international consensus that future human activities shall be considered in safety assessments of deep geological repositories. Based on generally accepted principles and the Swedish Radiation Safety Authority's, SSM's, regulations SSM FS 2008:21 and SSM FS 2008:37, the future human actions considered in this part of the safety assessment are restricted to global pollution and actions that: - are carried out after the sealing of the repository, - take place at or close to the repository site, - are unintentional, i.e. are carried out when the location of the repository is unknown, its purpose forgotten or the consequences of the action are unknown, - impair the safety functions of the repository's barriers. However, in line with SSM's general guidance /SSM 2008a/, future human actions and their impact on the repository are evaluated separately, and are not included in the main scenario reference evolution or in the risk summation

Handling of future human actions in the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

authorities in their review of SR-Can /Dverstorp and Stroemberg 2008/ maintain that the state, rather than SKB, is expected to be responsible for the supervision and monitoring of the repository after sealing. Man is dependent on, and influences, the environment in which he lives. After the repository has been closed, future generations should be able to utilise the repository site according to their needs without jeopardising their health. In the case of a final repository of the KBS-3 type, there are, however, inevitably examples of activities that, if carried out carelessly or without knowledge of the repository, could result in exposure to radiotoxic elements from the spent fuel. Therefore, there is an international consensus that future human activities shall be considered in safety assessments of deep geological repositories. Based on generally accepted principles and the Swedish Radiation Safety Authority's, SSM's, regulations SSM FS 2008:21 and SSM FS 2008:37, the future human actions considered in this part of the safety assessment are restricted to global pollution and actions that: - are carried out after the sealing of the repository, - take place at or close to the repository site, - are unintentional, i.e. are carried out when the location of the repository is unknown, its purpose forgotten or the consequences of the action are unknown, - impair the safety functions of the repository's barriers. However, in line with SSM's general guidance /SSM 2008a/, future human actions and their impact on the repository are evaluated separately, and are not included in the main scenario reference evolution or in the risk summation

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-12-01

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

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

International Nuclear Information System (INIS)

2000-12-01

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

Independent Calculations for the SR Can Assessment. External review contribution in support of SKI's and SSI's review of SR-Can

Energy Technology Data Exchange (ETDEWEB)

Maul, Philip; Robinson, Peter; Bond, Alex; Benbow, Steven (Quintessa Limited, Henley-on-Thames (GB))

2008-03-15

SKB has published the SR-Can assessment of a deep repository for spent fuel at either the Forsmark or Laxemar sites. This is the final assessment prior to a formal regulatory submission. A number of independent calculations have been undertaken in support of SKI's review of SR-Can. The types of calculations are: 1. direct checks of specified SKB calculations; 2. reproduction of SKB computer calculations with independent codes, to ensure that what SKB has done is properly understood, and to check that the calculations are properly documented; and 3. independent calculations to investigate particular aspects of the safety case. The data used by SKB in its Performance Assessment calculations have not been subject to detailed review. The independent calculations provide information on: 1. where independent calculations have been able to provide support for the arguments put forward by SKB; 2. areas where insufficient information has been provided by SKB to enable a third party to reproduce the SR-Can calculations; and 3. areas where calculations lead to questions about the validity of SKB's arguments. The timescale for the production of the present report has been determined by the timescales for SKI's review of the SR-Can assessment. As a result, some of the independent calculations referred to have not been fully documented, and this will be carried out in 2008. The following conclusions have been drawn. 1. SKB has worked hard to respond to criticisms of previous performance assessments, and SR-Can is an impressive piece of work. 2. In several areas either insufficient or inconsistent information has been presented so that a full reproduction of SKB's calculations has not been possible. This is an important area where SKB will need to improve the presentation of its assessment for SR-Site. 3. There are several areas where SKB's description of post-closure repository evolution needs to be further reviewed. Overall SKB have given only limited

Review comments on the SR 97 post-closure safety assessment

International Nuclear Information System (INIS)

Geier, J.

2000-01-01

These review comments concern an assessment of the long-term safety of a deep repository for spent nuclear fuel, titled Safety Report 97 (SR 97), which was prepared by the Swedish Nuclear Fuel Waste Management Company (SKB). The primary focus of this review is on hydrogeologic issues relating to groundwater flow, hydrologic uncertainty, and the potential for radionuclide transport from leaking canisters. The main hydrological model that was used in SR 97 is based on a continuum conceptual model of groundwater flow in fractured bedrock. Major problems with this model include the following: The validity of the continuum model is arguable for the type of rock that is present at these sites. The suitability of the model for the intended purpose of predicting streamlines and travel times for groundwater flow through the rock mass has not been adequately demonstrated. The comparison with alternative, discrete models yielded more divergent results than has been recognized in the SR 97 reports. The comparison with alternative models did not consider significant, realistic sources of uncertainty in the alternative models, evaluation of which would have likely led to greater divergence. The SR 97 model of radionuclide transport is based on a 1-D streamtube formulation, within which the predicted release of radionuclides to the biosphere is dominated by a parameter called the F ratio. A key factor in this parameter is the flow wetted surface. All of the hydrologic models used in SR 97 relied upon essentially the same set of geometric assumptions to estimate flow wetted surface from conductive fracture frequency in boreholes. Hence the predictions of the alternative models are not independent. Alternative methods of estimating flow wetted surface are needed to obtain a realistic evaluation of the uncertainty regarding radionuclide release. The alternative 3-D hydrologic models were used only to predict streamtube parameters, not for actual transport simulations. Hence the

Groundwater flow and radionuclide transport modelling using CONNECTFLOW in support of the SR Can assessment

International Nuclear Information System (INIS)

Hartley, Lee; Cox, Ian; Holton, David; Hunter, Fiona; Joyce, Steve; Gylling, Bjoern; Lindgren, Maria

2004-09-01

SKB is currently pursuing site investigations for a deep repository in the municipalities of Oesthammar and Oskarshamn. The investigations are conducted in two stages; an initial phase followed by a complete site investigation phase. The favoured alternative for the location of the encapsulation plant is at Oskarshamn, where it would operate in conjunction with the existing interim storage facility. These two planning applications will each require a report on the long-term safety of the deep repository. In the case of the encapsulation plant, such a report will demonstrate that a repository for the sealed canisters will meet the requirements on long-term safety set up by the Swedish authorities. The two safety reports will be referred to as SR-Can and SR-Site, for the encapsulation plant and repository, respectively. SR-Can will be based on site data from the initial site investigation phase and SR-Site on data from the complete site investigation. The preliminary safety evaluations for each site will be carried out as sub-tasks within the SR-Can project. The main purposes of those evaluations are to: Determine whether earlier judgements of the suitability of the candidate area for a deep repository with respect to long-term safety holds up in the light of borehole data; Provide feed-back to continued site investigations and site-specific repository design. A proposed methodology for the SR-Can assessment has been published in SKB TR-03-08. The methodology envisaged the use of both continuum porous medium (CPM) and discrete fracture network (DFN) models on a range of scales to investigate the groundwater flow and radionuclide transport from a deep disposal facility to the biosphere. The modelling must address the effects of variable groundwater density and transients. Transients occur naturally as a consequence of changes in climate states (temperate, periglacial and glacial) and during the operational and immediate post-closure phases of the repository. Key

Corrosion calculations report for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

This report is a compilation of the quantitative assessments of corrosion of the copper canisters in a KBS-3 repository. The calculations are part of the safety assessment SR-Site that is the long-term safety assessment to support the license application for building a final repository for spent nuclear fuel at Forsmark, Sweden. The safety assessment methodology gives the frame for the structured and documented approach to assess all conceivable corrosion processes. The quantitative assessments are done in different ways depending on the nature of the process and on the implications for the long-term safety. The starting point for the handling of the corrosion processes is the description of all known corrosion processes for copper with the current knowledge base and applied to the specific system and geology. Already at this stage some processes are excluded for further analysis, for example if the repository environment is not a sufficient prerequisite for the process to occur. The next step is to identify processes where the extent of corrosion could be bounded, e.g. by a mass balance approach. For processes where a mass balance is not limiting, the mass transport of corrodants (or corrosion products) is taken into account. A simple approach would be just to calculate the diffusive transport of corrodants through the bentonite, but generally the transport resistance for the interface between groundwater in a rock fracture intersecting the deposition hole and the bentonite buffer is more important. In SR-Site, the concept of equivalent flowrate, Q eq , is used. This assessment is done integrated with the evaluation of the geochemical and hydrogeological evolution of the repository. For most of the corrosion processes analysed, the corrosion depth is much smaller than the copper shell thickness, even for the assessment time of 10 6 years. Several processes give corrosion depths less than 100 μm, but no process give corrosion depths larger than a few millimetres

Corrosion calculations report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This report is a compilation of the quantitative assessments of corrosion of the copper canisters in a KBS-3 repository. The calculations are part of the safety assessment SR-Site that is the long-term safety assessment to support the license application for building a final repository for spent nuclear fuel at Forsmark, Sweden. The safety assessment methodology gives the frame for the structured and documented approach to assess all conceivable corrosion processes. The quantitative assessments are done in different ways depending on the nature of the process and on the implications for the long-term safety. The starting point for the handling of the corrosion processes is the description of all known corrosion processes for copper with the current knowledge base and applied to the specific system and geology. Already at this stage some processes are excluded for further analysis, for example if the repository environment is not a sufficient prerequisite for the process to occur. The next step is to identify processes where the extent of corrosion could be bounded, e.g. by a mass balance approach. For processes where a mass balance is not limiting, the mass transport of corrodants (or corrosion products) is taken into account. A simple approach would be just to calculate the diffusive transport of corrodants through the bentonite, but generally the transport resistance for the interface between groundwater in a rock fracture intersecting the deposition hole and the bentonite buffer is more important. In SR-Site, the concept of equivalent flowrate, Q{sub eq}, is used. This assessment is done integrated with the evaluation of the geochemical and hydrogeological evolution of the repository. For most of the corrosion processes analysed, the corrosion depth is much smaller than the copper shell thickness, even for the assessment time of 106 years. Several processes give corrosion depths less than 100 mum, but no process give corrosion depths larger than a few

Review of SKB's interim report of SR-Can: SKI's and SSI's evaluation of SKB's up-dated methodology for safety assessment

Energy Technology Data Exchange (ETDEWEB)

Dverstorp, Bjoern; Moberg, Leif; Wiebert, Anders; Xu Shulan [Swedish Radiation Protection Authority, Stockholm (Sweden); Stroemberg, Bo; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Sundstroem, Benny; Toverud, Oeivind [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

2005-07-01

This report presents the findings of a review of the Swedish Nuclear Fuel and Waste Management Co.'s (SKB) interim report of the safety assessment SR-Can (SKB TR 04-11), conducted by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). SKB's interim report describes and exemplifies the safety assessment methodology that SKB plans to use in the oncoming licence applications for an encapsulation plant and a final repository for spent nuclear fuel. The authorities' review takes into account the findings of an international peer review of SKB's interim report. The authorities conclude that SKB has improved its safety assessment methodology in several aspects compared to earlier safety reports. Among other things the authorities commend SKB for giving a comprehensive account of relevant regulations and guidance, and for the systematic approach to identification and documentation of features, events and processes that need to be considered in the safety assessment. However, the authorities also conclude that important parts of SKB's method need to be further developed before they are mature enough to be used as a basis for a license application. The authorities' overall assessment is summarised in chapter 8 of this report.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-12-01

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

The biosphere at Laxemar. Data, assumptions and models used in the SR-Can assessment

Energy Technology Data Exchange (ETDEWEB)

Karlsson, Sara; Kautsky, Ulrik; Loefgren, Anders; Soederbaeck, Bjoern [eds.

2006-10-15

This is essentially a compilation of a variety of reports concerning the site investigations, the research activities and information derived from other sources important for the safety assessment. The main objective is to present prerequisites, methods and data used, in the biosphere modelling for the safety assessment SR-Can at the Laxemar site. A major part of the report focuses on how site-specific data are used, recalculated or modified in order to be applicable in the safety assessment context; and the methods and sub-models that are the basis for the biosphere modelling. Furthermore, the assumptions made as to the future states of surface ecosystems are mainly presented in this report. A similar report is provided for the Forsmark area. This report summarises the method adopted for safety assessment following a radionuclide release into the biosphere. The approach utilises the information about the site as far as possible and presents a way of calculating risk to humans. A central tool in the work is the description of the topography, where there is good understanding of the present conditions and the development over time is fairly predictable. The topography affects surface hydrology, sedimentation, size of drainage areas and the characteristics of ecosystems. Other parameters are human nutritional intake, which is assumed to be constant over time, and primary production (photosynthesis), which also is a fairly constant parameter over time. The Landscape Dose Factor approach (LDF) gives an integrated measure for the site and also resolves the issues relating to the size of the group with highest exposure. If this approach is widely accepted as method, still some improvements and refinement are necessary in collecting missing site data, reanalysing site data, reviewing radionuclide specific data, reformulating ecosystem models and evaluating the results with further sensitivity analysis.

The biosphere at Laxemar. Data, assumptions and models used in the SR-Can assessment

International Nuclear Information System (INIS)

Karlsson, Sara; Kautsky, Ulrik; Loefgren, Anders; Soederbaeck, Bjoern

2006-10-01

This is essentially a compilation of a variety of reports concerning the site investigations, the research activities and information derived from other sources important for the safety assessment. The main objective is to present prerequisites, methods and data used, in the biosphere modelling for the safety assessment SR-Can at the Laxemar site. A major part of the report focuses on how site-specific data are used, recalculated or modified in order to be applicable in the safety assessment context; and the methods and sub-models that are the basis for the biosphere modelling. Furthermore, the assumptions made as to the future states of surface ecosystems are mainly presented in this report. A similar report is provided for the Forsmark area. This report summarises the method adopted for safety assessment following a radionuclide release into the biosphere. The approach utilises the information about the site as far as possible and presents a way of calculating risk to humans. A central tool in the work is the description of the topography, where there is good understanding of the present conditions and the development over time is fairly predictable. The topography affects surface hydrology, sedimentation, size of drainage areas and the characteristics of ecosystems. Other parameters are human nutritional intake, which is assumed to be constant over time, and primary production (photosynthesis), which also is a fairly constant parameter over time. The Landscape Dose Factor approach (LDF) gives an integrated measure for the site and also resolves the issues relating to the size of the group with highest exposure. If this approach is widely accepted as method, still some improvements and refinement are necessary in collecting missing site data, reanalysing site data, reviewing radionuclide specific data, reformulating ecosystem models and evaluating the results with further sensitivity analysis

International Peer Review of Swedish Nuclear Fuel and Waste Management Company's SR-Can interim report

International Nuclear Information System (INIS)

Sagar, Budhi; Bailey, Lucy; Bennett, David G.; Egan, Mike; Roehlig, Klaus

2004-12-01

SKB has produced an interim safety assessment report as part of its work to develop a licence application for the construction of a spent nuclear fuel encapsulation plant. The purpose of the interim report is to set out and demonstrate SKB's proposed methodology for long-term safety assessment. The aim of producing an interim report is to allow the Swedish regulatory authorities (SKI and SSI) to review and comment on SKB's proposed methodology before it is used in support of a formal licence application. To help inform their review of SKB's proposed methodology, the authorities appointed an international review team (IRT) to carry out a review of SKB's interim safety assessment report. Comments from the IRT are presented in this document and will be considered by the regulatory authorities in developing their own view of SKB's proposed methodology. The IRT's review included examination of SKB's documentation (the 'Interim Main Report of the Safety Assessment SR-Can' and four supporting documents) and hearings with SKB staff and contractors. The hearings provided an opportunity for the IRT to discuss the SR-Can safety assessment with the authors and contributors to SKB's work. As directed by SKI and SSI, the IRT's review focused on methodological aspects and sought to determine whether SKB's proposed safety assessment methodology: (i) is fit for the purpose of supporting a licence application; (ii) has a reasonable prospect of leading to a safety assessment that is sufficiently comprehensive, reproducible, traceable and transparent; (iii) is compatible with the authorities' regulations and guidance. No evaluation of long term safety or site acceptability was attempted by the IRT. At the request of SKI and SSI, the IRT's review considered and made recommendations on the following issues: Description of the initial state of the repository and its components; Description of features, events and processes (FEPs) relevant to repository evolution; Strategy for safety

A 3S Risk ?3SR? Assessment Approach for Nuclear Power: Safety Security and Safeguards.

Energy Technology Data Exchange (ETDEWEB)

Forrest, Robert; Reinhardt, Jason Christian; Wheeler, Timothy A.; Williams, Adam David

2017-11-01

Safety-focused risk analysis and assessment approaches struggle to adequately include malicious, deliberate acts against the nuclear power industry's fissile and waste material, infrastructure, and facilities. Further, existing methods do not adequately address non- proliferation issues. Treating safety, security, and safeguards concerns independently is inefficient because, at best, it may not take explicit advantage of measures that provide benefits against multiple risk domains, and, at worst, it may lead to implementations that increase overall risk due to incompatibilities. What is needed is an integrated safety, security and safeguards risk (or "3SR") framework for describing and assessing nuclear power risks that can enable direct trade-offs and interactions in order to inform risk management processes -- a potential paradigm shift in risk analysis and management. These proceedings of the Sandia ePRA Workshop (held August 22-23, 2017) are an attempt to begin the discussions and deliberations to extend and augment safety focused risk assessment approaches to include security concerns and begin moving towards a 3S Risk approach. Safeguards concerns were not included in this initial workshop and are left to future efforts. This workshop focused on four themes in order to begin building out a the safety and security portions of the 3S Risk toolkit: 1. Historical Approaches and Tools 2. Current Challenges 3. Modern Approaches 4. Paths Forward and Next Steps This report is organized along the four areas described above, and concludes with a summary of key points. 2 Contact: rforres@sandia.gov; +1 (925) 294-2728

Nuclear safety activities in SR Slovenia in 1985

International Nuclear Information System (INIS)

1986-09-01

Currently Yugoslavia has one 632 MWe nuclear power plant of PWR design, located at Krsko in the Socialist Republic of Slovenia. NPP Krsko, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in SR Slovenia are mostly related to upgrading the safety of our NPP Krsko and to develop capabilities to be used for the future units. This report presents safety related organizations in SR Slovenia and their activities performed in 1985. (author)

Nuclear safety activities in SR Slovenia in 1985

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-09-15

Currently Yugoslavia has one 632 MWe nuclear power plant of PWR design, located at Krsko in the Socialist Republic of Slovenia. NPP Krsko, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in SR Slovenia are mostly related to upgrading the safety of our NPP Krsko and to develop capabilities to be used for the future units. This report presents safety related organizations in SR Slovenia and their activities performed in 1985. (author)

Nuclear safety activities in the SR of Slovenia in 1986

Energy Technology Data Exchange (ETDEWEB)

Susnik, J [Inst. Jozef Stefan, Ljubljana (Slovenia)

1987-06-15

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1986. (author)

Nuclear safety activities in the SR of Slovenia in 1986

International Nuclear Information System (INIS)

Susnik, J.

1987-06-01

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1986. (author)

Data and uncertainty assessment for radionuclide Kd partitioning coefficients in granitic rock for use in SR-Can calculations

International Nuclear Information System (INIS)

Crawford, James; Neretnieks, Ivars; Malmstroem, Maria

2006-10-01

SKB is currently preparing licence applications related to the proposed deep repository for spent nuclear fuel as well as the encapsulation plant required for canister fabrication. The present report is one of several specific data reports that form the data input to an interim safety report (SR-Can) for the encapsulation plant licence application. This report concerns the derivation and recommendation of generic K d data (i.e. linear partitioning coefficients) for safety assessment modelling of far-field radionuclide transport in fractured granitic rock. The data are derived for typical Swedish groundwater conditions and rock types distinctive of those found on the Simpevarp peninsula and Forsmark. Data have been derived for 8 main elements (Cs, Sr, Ra, Ni, Th, U, Np, Am) and various oxidation states. The data have also been supplied with tentative correction factors to account for artefacts that have not been previously considered in detail in previous compilations. For the main reviewed solutes the data are given in terms of a best estimate K d value assumed to be the median of the aggregate set of selected data. A range corresponding to the 25-75% inter-quartile range is also specified and probable ranges of uncertainty are estimated in the form of an upper and lower limit to the expected variability. Data for an additional 19 elements that have not been reviewed are taken from a previous compilation by Carbol and Engkvist

International Expert Review of SRCan: Engineered Barrier Issues. External review contribution in support of SKI's and SSI's review of SR-Can

International Nuclear Information System (INIS)

Savage, David; Bennett, David; Apted, Mick; Saellfors, Goeran; Saario, Timo; Segle, Peter

2008-03-01

The Swedish Nuclear Fuel and Waste Management Company (SKB) has recently submitted a license application for the construction of a spent fuel encapsulation plant. SKB plans to submit a further license application in 2009 for the construction of a repository for the disposal spent nuclear fuel. In connection with the first of these applications, SKB published a safety report, known as SR-Can, which assessed the safety of a spent-fuel repository. The Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Authority (SSI) (the Authorities) will make formal reviews of the licence applications, and have, therefore, jointly commissioned a team of independent experts to assess and provide comments on SKB's safety reports. The Authorities will consider the views of the independent review team in completing their own reviews. This document presents the comments and findings of the Engineered Barrier System (EBS) review group on SR-Can. The SR-Can safety report includes an examination of EBS design and performance for a range of scenarios, including expected repository evolution and possible variant scenarios, that together address processes and events that might result in the loss of certain repository safety functions. Furthermore, a series of sensitivity analyses is also presented that provides helpful insights into the relative importance of many key parameters and processes related to the EBS. In general, the explanatory text of the SR-Can safety report is clear, and the cited references provide adequate technical justifications for the assumptions, models, and data that are abstracted into the SR-Can safety report. The review group considers, therefore, that SKB's development of SR-Can has been a very valuable exercise, and that SKB should be congratulated on the breadth, depth and general clarity of its research and development and safety assessment programmes. Notwithstanding these successes, the EBS review group has identified a range of

International Expert Review of SRCan: Engineered Barrier Issues. External review contribution in support of SKI's and SSI's review of SR-Can

Energy Technology Data Exchange (ETDEWEB)

Savage, David (Quintessa Limited, Henley-on-Thames (GB)); Bennett, David (TerraSalus Limited, Oakham (GB)); Apted, Mick (Monitor Scientific LLC, Denver, CO (US)); Saellfors, Goeran (Chalmers Univ. of Technology, Goeteborg (SE)); Saario, Timo (VTT Materials and Building (FI)); Segle, Peter (Inspecta, Stockholm (SE))

2008-03-15

The Swedish Nuclear Fuel and Waste Management Company (SKB) has recently submitted a license application for the construction of a spent fuel encapsulation plant. SKB plans to submit a further license application in 2009 for the construction of a repository for the disposal spent nuclear fuel. In connection with the first of these applications, SKB published a safety report, known as SR-Can, which assessed the safety of a spent-fuel repository. The Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Authority (SSI) (the Authorities) will make formal reviews of the licence applications, and have, therefore, jointly commissioned a team of independent experts to assess and provide comments on SKB's safety reports. The Authorities will consider the views of the independent review team in completing their own reviews. This document presents the comments and findings of the Engineered Barrier System (EBS) review group on SR-Can. The SR-Can safety report includes an examination of EBS design and performance for a range of scenarios, including expected repository evolution and possible variant scenarios, that together address processes and events that might result in the loss of certain repository safety functions. Furthermore, a series of sensitivity analyses is also presented that provides helpful insights into the relative importance of many key parameters and processes related to the EBS. In general, the explanatory text of the SR-Can safety report is clear, and the cited references provide adequate technical justifications for the assumptions, models, and data that are abstracted into the SR-Can safety report. The review group considers, therefore, that SKB's development of SR-Can has been a very valuable exercise, and that SKB should be congratulated on the breadth, depth and general clarity of its research and development and safety assessment programmes. Notwithstanding these successes, the EBS review group has identified a range

Data and uncertainty assessment for radionuclide K{sub d} partitioning coefficients in granitic rock for use in SR-Can calculations

Energy Technology Data Exchange (ETDEWEB)

Crawford, James; Neretnieks, Ivars; Malmstroem, Maria [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Chemical Engineering and Technology

2006-10-15

SKB is currently preparing licence applications related to the proposed deep repository for spent nuclear fuel as well as the encapsulation plant required for canister fabrication. The present report is one of several specific data reports that form the data input to an interim safety report (SR-Can) for the encapsulation plant licence application. This report concerns the derivation and recommendation of generic K{sub d} data (i.e. linear partitioning coefficients) for safety assessment modelling of far-field radionuclide transport in fractured granitic rock. The data are derived for typical Swedish groundwater conditions and rock types distinctive of those found on the Simpevarp peninsula and Forsmark. Data have been derived for 8 main elements (Cs, Sr, Ra, Ni, Th, U, Np, Am) and various oxidation states. The data have also been supplied with tentative correction factors to account for artefacts that have not been previously considered in detail in previous compilations. For the main reviewed solutes the data are given in terms of a best estimate K{sub d} value assumed to be the median of the aggregate set of selected data. A range corresponding to the 25-75% inter-quartile range is also specified and probable ranges of uncertainty are estimated in the form of an upper and lower limit to the expected variability. Data for an additional 19 elements that have not been reviewed are taken from a previous compilation by Carbol and Engkvist.

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

International Nuclear Information System (INIS)

Karlsson, Sara; Bergstroem, Ulla

2002-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-05-01

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

T-H-M couplings in rock. Overview of results of importance to the SR-Can safety assessment

International Nuclear Information System (INIS)

Hoekmark, Harald; Faelth, Billy; Wallroth, Thomas

2006-09-01

This report deals with THM processes in rock hosting a KBS-nuclear waste repository. The issues addressed are mechanically and thermo-mechanically induced changes of the hydraulic conditions in the near-field and in the far-field, and the risk of stress-induced failure, spalling, in the walls of deposition holes. These changes are examined for the construction and operational phases, the initial temperate period and a subsequent glacial cycle. The report was compiled to be used as reference and background for corresponding parts of the safety report SR-Can. The near-field is analyzed using thermo-mechanical DEC models. There are a number of models for each of the three sites Forsmark, Simpevarp and Laxemar. Parameter values of intact rock and rock mass mechanical and thermo-mechanical properties were obtained from the site descriptive models. Layout data, i.e. the number of canisters, the geometry of the repository openings, the tunnel spacing and the canister spacing are in accordance with rules and guidelines given in general design- and layout documents. Heat generation data, i.e. the initial canister power and the canister power decay are in accordance with data given in the SR-Can main report. The effects of changes in nearfield stresses during the different phases of the repository's lifetime are evaluated by comparing numerically obtained stresses and deformations on a number of explicitly modelled near-field fractures with empirical and theoretical stress-transmissivity laws and with empirically based slip-transmissivity estimates. For the near-field it is concluded that substantial transmissivity increases are found only very close to the repository openings. Bounding estimates, judged to be valid for the entire load sequence, are made of the extent and magnitude of the hydraulic disturbance. At distances larger than about 1.5 m from the tunnel periphery, transmissivity increases are concluded to be too small and unsystematic to be of any concern. The

T-H-M couplings in rock. Overview of results of importance to the SR-Can safety assessment

Energy Technology Data Exchange (ETDEWEB)

Hoekmark, Harald; Faelth, Billy [Clay Technology AB, Lund (Sweden); Wallroth, Thomas [BERGAB, Goeteborg (Sweden)

2006-09-15

This report deals with THM processes in rock hosting a KBS-nuclear waste repository. The issues addressed are mechanically and thermo-mechanically induced changes of the hydraulic conditions in the near-field and in the far-field, and the risk of stress-induced failure, spalling, in the walls of deposition holes. These changes are examined for the construction and operational phases, the initial temperate period and a subsequent glacial cycle. The report was compiled to be used as reference and background for corresponding parts of the safety report SR-Can. The near-field is analyzed using thermo-mechanical DEC models. There are a number of models for each of the three sites Forsmark, Simpevarp and Laxemar. Parameter values of intact rock and rock mass mechanical and thermo-mechanical properties were obtained from the site descriptive models. Layout data, i.e. the number of canisters, the geometry of the repository openings, the tunnel spacing and the canister spacing are in accordance with rules and guidelines given in general design- and layout documents. Heat generation data, i.e. the initial canister power and the canister power decay are in accordance with data given in the SR-Can main report. The effects of changes in nearfield stresses during the different phases of the repository's lifetime are evaluated by comparing numerically obtained stresses and deformations on a number of explicitly modelled near-field fractures with empirical and theoretical stress-transmissivity laws and with empirically based slip-transmissivity estimates. For the near-field it is concluded that substantial transmissivity increases are found only very close to the repository openings. Bounding estimates, judged to be valid for the entire load sequence, are made of the extent and magnitude of the hydraulic disturbance. At distances larger than about 1.5 m from the tunnel periphery, transmissivity increases are concluded to be too small and unsystematic to be of any concern. The

Biosphere analyses for the safety assessment SR-Site - synthesis and summary of results

International Nuclear Information System (INIS)

Saetre, Peter

2010-12-01

ice free period between two glaciations. The radionuclide model used in SR-Site has been improved in several important ways since previous safety assessments conducted by SKB. For example, the aquatic and terrestrial ecosystems are handled in the same model, which gives a continuous transition from the sea stage to the lake and terrestrial stages. Transport and accumulation in till (lower regolith) is represented in the model. The uptake by plants is included in the mass-balance, and it is related to biomass growth. Moreover, parameter values including hydrological flows, sedimentation and resuspension rates, biomass growth rates, gas exchange rates, as well as element specific distribution coefficients and concentration rations, were as far as possible based on site data. One endpoint from the simulations with the radionuclide model was the landscape dose conversion factors (LDFs). The LDF represents the mean annual effective dose over lifetime for an individual living in the most contaminated area, assuming a constant unit release rate (1Bq/y). In the safety assessment, the maximum LDF for each nuclide have been selected from the biosphere object at the time yielding the highest unit release dose, and consequently LDFs from different nuclides does not necessarily match the same group of exposed individuals with respect to point in time or location in the landscape. In the SR-Site main report, the resulting dose is presented when the maximum LDF is multiplied by a release. The potential effect of a radionuclide release on non-human biota in Forsmark is also assessed

Determination and assessment of the concentration limits to be used in SR-Can. Supplement to TR-06-32

International Nuclear Information System (INIS)

Grive, Mireia; Domenech, Cristina; Montoya, Vanessa; Garcia, David; Duro, Lara

2010-09-01

This document complements and updates the report TR-06-32, Determination and assessment of the concentration limits to be used in SR-Can, in which the solubility limits of different radionuclides in the near field system and under the different scenarios selected by SKB were assessed. Since 2006, several important changes in different fields affecting solubility assessment calculations have been reported. These changes basically concern some of the thermodynamic data used in the calculations and the groundwater compositions for scenarios of interest defined by SKB. In this document we update the thermodynamic data corresponding to Ni, Zr, Th and U and we describe the thermodynamic database selected for Pb. This document also reports the update of the assessment of the concentration limits to be used in SR-Can, which has been done considering the recent thermodynamic database updates and the new groundwater compositions of interest supplied by SKB. Finally, we also present the Simple Functions spreadsheet tool, born from the need of having a confident and easy-to-handle tool to calculate solubility limits of some radionuclides under determined conditions in an agile and relatively fast manner

Determination and assessment of the concentration limits to be used in SR-Can. Supplement to TR-06-32

Energy Technology Data Exchange (ETDEWEB)

Grive, Mireia; Domenech, Cristina; Montoya, Vanessa; Garcia, David; Duro, Lara

2010-09-15

This document complements and updates the report TR-06-32, Determination and assessment of the concentration limits to be used in SR-Can, in which the solubility limits of different radionuclides in the near field system and under the different scenarios selected by SKB were assessed. Since 2006, several important changes in different fields affecting solubility assessment calculations have been reported. These changes basically concern some of the thermodynamic data used in the calculations and the groundwater compositions for scenarios of interest defined by SKB. In this document we update the thermodynamic data corresponding to Ni, Zr, Th and U and we describe the thermodynamic database selected for Pb. This document also reports the update of the assessment of the concentration limits to be used in SR-Can, which has been done considering the recent thermodynamic database updates and the new groundwater compositions of interest supplied by SKB. Finally, we also present the Simple Functions spreadsheet tool, born from the need of having a confident and easy-to-handle tool to calculate solubility limits of some radionuclides under determined conditions in an agile and relatively fast manner

Determination and assessment of the concentration limits to be used in SR-Can

Energy Technology Data Exchange (ETDEWEB)

Duro, L; Grive, M; Cera, E; Gaona, X; Domenech, C; Bruno, J [Enviros Spain S.L., Barcelona (Spain)

2006-12-15

This report presents the results for solubility limit calculations for the SR-Can assessment. It has been organized into five chapters that constitute the core of the report, supported by several appendices containing additional and supporting information. The updated thermodynamic database used to conduct the solubility calculations has been issued as a separate report. The near field system for which the concentration limits of the radionuclides are assessed and the scenarios selected by SKB to calculate the solubility limits are thoroughly described. Several sources of information have been used to support the calculated solubility limits. In particular results from selected spent fuel dissolution experiments and natural analogue data are discussed to introduce the proper perspective to the results from the thermodynamic calculations. In addition, the main conceptual and numerical uncertainties associated to the assessment of the concentration limits of each element are numerically evaluated and discussed. Equilibrium calculations have been conducted to select the solubility limiting solid phase for each element. Furthermore a sensitivity analysis of parameters of interest for each element is presented and the impact of the uncertainties identified on the solubility of each element quantified. The results are presented in a series of tables containing the calculated solubility for each radionuclide under the reference conditions. Finally concentration limits that are recommended result from the expert judgement built-up around the various sources of information together with the quantification of radionuclide solubility data and their associated uncertainties. The results are compared to previous solubility limits determination performed by SKB in SR 97, as well as the recommended values from other HLNW management organisations.

Determination and assessment of the concentration limits to be used in SR-Can

International Nuclear Information System (INIS)

Duro, L.; Grive, M.; Cera, E.; Gaona, X.; Domenech, C.; Bruno, J.

2006-12-01

This report presents the results for solubility limit calculations for the SR-Can assessment. It has been organized into five chapters that constitute the core of the report, supported by several appendices containing additional and supporting information. The updated thermodynamic database used to conduct the solubility calculations has been issued as a separate report. The near field system for which the concentration limits of the radionuclides are assessed and the scenarios selected by SKB to calculate the solubility limits are thoroughly described. Several sources of information have been used to support the calculated solubility limits. In particular results from selected spent fuel dissolution experiments and natural analogue data are discussed to introduce the proper perspective to the results from the thermodynamic calculations. In addition, the main conceptual and numerical uncertainties associated to the assessment of the concentration limits of each element are numerically evaluated and discussed. Equilibrium calculations have been conducted to select the solubility limiting solid phase for each element. Furthermore a sensitivity analysis of parameters of interest for each element is presented and the impact of the uncertainties identified on the solubility of each element quantified. The results are presented in a series of tables containing the calculated solubility for each radionuclide under the reference conditions. Finally concentration limits that are recommended result from the expert judgement built-up around the various sources of information together with the quantification of radionuclide solubility data and their associated uncertainties. The results are compared to previous solubility limits determination performed by SKB in SR 97, as well as the recommended values from other HLNW management organisations

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part III

International Nuclear Information System (INIS)

2011-01-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part I

International Nuclear Information System (INIS)

2011-01-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Bedrock Kd data and uncertainty assessment for application in SR-Site geosphere transport calculations

International Nuclear Information System (INIS)

Crawford, James

2010-12-01

The safety assessment SR-Site is undertaken to assess the safety of a potential geologic repository for spent nuclear fuel at the Forsmark and Laxemar sites. The present report is one of several reports that form the data input to SR-Site and contains a compilation of recommended K d data (i.e. linear partitioning coefficients) for safety assessment modelling of geosphere radionuclide transport. The data are derived for rock types and groundwater compositions distinctive of the site investigation areas at Forsmark and Laxemar. Data have been derived for all elements and redox states considered of importance for far-field dose estimates as described in /SKB 2010d/. The K d data are given in the form of lognormal distributions characterised by a mean (μ) and standard deviation (σ). Upper and lower limits for the uncertainty range of the recommended data are defined by the 2.5% and 97.5% percentiles of the empirical data sets. The best estimate K d value for use in deterministic calculations is given as the median of the K d distribution

International Expert Review of SRCan: Site Investigation Aspects. External review contribution in support of SKI's and SSI's review of SR-Can. INSITE/OVERSITE

International Nuclear Information System (INIS)

2008-03-01

As a first evaluation of long-term safety for KBS-3 repositories at Forsmark and Laxemar, the SIG (Site Investigation Group) found SR-Can to be a well-produced and generally well-argued safety assessment. Overall, SKB is to be complimented on this project. Members of of the two groups INSITE and OVERSITE within the SIG had somewhat differing views on how well SKB had made use of the site data available at the end of the SDM 1.2 stage of investigations. This difference is less to do with the extent of site characterisation than of its use and application, reflecting the different levels of maturity of SKB's geosphere and biosphere assessment programmes. The more recent and current work on the sites means that our concerns expressed in this review should, to a large extent, be addressable in or prior to SR-Site, provided SKB is so minded. However, we acknowledge that some of the issues we raise will not be fully resolved until underground rock characterisation from excavations or longer records of surface conditions are available. There are also some key aspects of SKB's methodology still under development that would benefit from review prior to their use in SR-Site. More space in the currently pressing schedule would allow for this review and a consequent increase in confidence. In any case, the authorities should be aware that SKB may face residual programmatic risks, associated principally with the underground design and layout (and their knockon effects into performance), even after SR-Site. An early understanding of some of these relationships would be helped by a plan (at least on an outline level) of the underground characterisation programme. We also note that many engineering matters are still to be confronted, not least the EBS design and its implementation, along with the treatment of high stresses, if Forsmark is selected. However, our views on the nature of the SR-Can analysis and the way in which site data have been utilised in it (our principal remit

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part I; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del I

Energy Technology Data Exchange (ETDEWEB)

2011-07-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part II; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del II

Energy Technology Data Exchange (ETDEWEB)

2011-07-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part III; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del III

Energy Technology Data Exchange (ETDEWEB)

2011-07-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Bedrock K{sub d} data and uncertainty assessment for application in SR-Site geosphere transport calculations

Energy Technology Data Exchange (ETDEWEB)

Crawford, James (Kemakta Konsult AB, Stockholm (Sweden))

2010-12-15

The safety assessment SR-Site is undertaken to assess the safety of a potential geologic repository for spent nuclear fuel at the Forsmark and Laxemar sites. The present report is one of several reports that form the data input to SR-Site and contains a compilation of recommended K{sub d} data (i.e. linear partitioning coefficients) for safety assessment modelling of geosphere radionuclide transport. The data are derived for rock types and groundwater compositions distinctive of the site investigation areas at Forsmark and Laxemar. Data have been derived for all elements and redox states considered of importance for far-field dose estimates as described in /SKB 2010d/. The K{sub d} data are given in the form of lognormal distributions characterised by a mean (mu) and standard deviation (sigma). Upper and lower limits for the uncertainty range of the recommended data are defined by the 2.5% and 97.5% percentiles of the empirical data sets. The best estimate K{sub d} value for use in deterministic calculations is given as the median of the K{sub d} distribution

Development of Landscape Dose Factors for dose assessments in SR-Can

Energy Technology Data Exchange (ETDEWEB)

Avila, Rodolfo; Ekstroem, Per-Anders [Facilia AB, Bromma (Sweden); Kautsky, Ulrik [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

2006-08-15

In previous safety assessments Ecosystem Dose Factors (EDFs), were derived from estimates of doses to the most exposed group resulting from constant unit radionuclide release rates over 10,000 years to various ecosystem types, e.g. mires, agricultural lands, lakes and marine ecosystems. A number of limitations of the EDF approach have been identified. The objectives of this report is to further develop the EDF approach, in order to resolve the identified limitations, and to use the improved approach for deriving Dose Conversion Factors for use in the SR-Can risk assessments. The Dose Conversion Factors derived in this report are named Landscape Dose Factors (LDFs). It involves modelling the fate of the radionuclides in the whole landscape, which develops from a sea to a inland situation during 20,000 years. Both candidate sites studies in SR-Can, Forsmark and Laxemar, are included in the study. As a basis for the modelling, the period starting at the beginning of the last interglacial (8,000 BC) is used, over which releases from a hypothetical repository were assumed to take place. For the present temperate period, the overall development of the biosphere at each site is outlined in a 1,000 year perspective and beyond, essentially based on the ongoing shoreline displacement and the understanding on the impact this has on the biosphere. The past development, i.e. from deglaciation to the present time, is inferred from geological records and associated reconstructions of the shore-line. For each time step of 1,000 years, the landscape at the site is described as a number of interconnected biosphere objects constituting an integrated landscape model of each site. The water fluxes through the objects were estimated from the average run-off at the site, the areas of the objects and their associated catchment areas. Radionuclides in both dissolved and particulate forms were considered in the transport calculations. The transformation between ecosystems was modelled as

Adverse Outcome Pathways can drive non-animal approaches for safety assessment.

Science.gov (United States)

Burden, Natalie; Sewell, Fiona; Andersen, Melvin E; Boobis, Alan; Chipman, J Kevin; Cronin, Mark T D; Hutchinson, Thomas H; Kimber, Ian; Whelan, Maurice

2015-09-01

Adverse Outcome Pathways (AOPs) provide an opportunity to develop new and more accurate safety assessment processes for drugs and other chemicals, and may ultimately play an important role in regulatory decision making. Not only can the development and application of AOPs pave the way for the development of improved evidence-based approaches for hazard and risk assessment, there is also the promise of a significant impact on animal welfare, with a reduced reliance on animal-based methods. The establishment of a useable and coherent knowledge framework under which AOPs will be developed and applied has been a first critical step towards realizing this opportunity. This article explores how the development of AOPs under this framework, and their application in practice, could benefit the science and practice of safety assessment, while in parallel stimulating a move away from traditional methods towards an increased acceptance of non-animal approaches. We discuss here the key areas where current, and future initiatives should be focused to enable the translation of AOPs into routine chemical safety assessment, and lasting 3Rs benefits. © 2015 The Authors. Journal of Applied Toxicology published by John Wiley & Sons Ltd.

Biosphere analyses for the safety assessment SR-Site - synthesis and summary of results

Energy Technology Data Exchange (ETDEWEB)

Saetre, Peter [comp.

2010-12-15

ice free period between two glaciations. The radionuclide model used in SR-Site has been improved in several important ways since previous safety assessments conducted by SKB. For example, the aquatic and terrestrial ecosystems are handled in the same model, which gives a continuous transition from the sea stage to the lake and terrestrial stages. Transport and accumulation in till (lower regolith) is represented in the model. The uptake by plants is included in the mass-balance, and it is related to biomass growth. Moreover, parameter values including hydrological flows, sedimentation and resuspension rates, biomass growth rates, gas exchange rates, as well as element specific distribution coefficients and concentration rations, were as far as possible based on site data. One endpoint from the simulations with the radionuclide model was the landscape dose conversion factors (LDFs).

Mineralogy and geochemistry of rocks and fracture fillings from Forsmark and Oskarshamn: Compilation of data for SR-Can

Energy Technology Data Exchange (ETDEWEB)

Drake, Henrik; Sandstroem, Bjoern [Isochron GeoConsulting HB, Goeteborg (Sweden); Tullborg, Eva-Lena [Terralogica AB, Graabo (Sweden)

2006-11-15

This report is a compilation of the so far available data for the safety assessment SR-Can carried out by SKB. The data consists of mineralogy, geochemistry, porosity, density and redox properties for both dominating rock types and fracture fillings at the Forsmark and Oskarshamn candidate areas. In addition to the compilation of existing information, the aim has been to identify missing data and to clarify some conception of e.g. deformation zones. The objective of the report is to present the available data requested for the modelling of the chemical stability of the two sites. The report includes no interpretation of the data.

The ecosystem models used for dose assessments in SR-Can

Energy Technology Data Exchange (ETDEWEB)

Avila, Rodolfo [Facilia AB, Bromma (Sweden)

2006-11-15

chronic contamination. From the simulations for the different release cases, activity concentrations in water and soil are obtained and then multiplied with the aggregated transfer factors to obtain concentrations in food products. For terrestrial ecosystems, the aggregated transfer factors in Becquerel per Kilogram of edible carbon in the food are used to calculate the activity intake and from this the effective dose rate per unit release to an adult individual. For aquatic ecosystems, only doses from the ingestion of water (for lakes) and food (for sea and lakes) are considered, as previous assessments have shown that in these types of ecosystems other exposure pathways give a very low contribution to the total doses. A sensitivity analysis of the ecosystem models is presented in the report, identifying which parameters have the largest effect on the simulation endpoints of interest. The endpoints considered are the fraction of the release that is retained in the ecosystem, the activity concentrations in soil, water and sediments, and the total dose rates from external exposure, inhalation, and ingestion of water and food. These endpoints are evaluated at different times within the simulation and a sensitivity analysis using the Morris method is carried out. For some of the scenarios considered in SR-Can, the LDF concept is not applicable. One of these scenarios comprises the contamination of ground caused by inadvertent drilling into the repository. Doses which would arise for a family using this contaminated ground for housing and food production are estimated. The other scenario which is assessed separately is the release of C-14 and Rn-222 from the repository in gaseous form, entering the biosphere via soil as a diffuse source. Pathways considered are doses from ingestion of C-14 and from inhalation of C-14 and Rn-222 outdoors as well as indoors. For these scenarios, specific dose calculations were carried out. The methods applied for these calculations and the

The ecosystem models used for dose assessments in SR-Can

International Nuclear Information System (INIS)

Avila, Rodolfo

2006-11-01

chronic contamination. From the simulations for the different release cases, activity concentrations in water and soil are obtained and then multiplied with the aggregated transfer factors to obtain concentrations in food products. For terrestrial ecosystems, the aggregated transfer factors in Becquerel per Kilogram of edible carbon in the food are used to calculate the activity intake and from this the effective dose rate per unit release to an adult individual. For aquatic ecosystems, only doses from the ingestion of water (for lakes) and food (for sea and lakes) are considered, as previous assessments have shown that in these types of ecosystems other exposure pathways give a very low contribution to the total doses. A sensitivity analysis of the ecosystem models is presented in the report, identifying which parameters have the largest effect on the simulation endpoints of interest. The endpoints considered are the fraction of the release that is retained in the ecosystem, the activity concentrations in soil, water and sediments, and the total dose rates from external exposure, inhalation, and ingestion of water and food. These endpoints are evaluated at different times within the simulation and a sensitivity analysis using the Morris method is carried out. For some of the scenarios considered in SR-Can, the LDF concept is not applicable. One of these scenarios comprises the contamination of ground caused by inadvertent drilling into the repository. Doses which would arise for a family using this contaminated ground for housing and food production are estimated. The other scenario which is assessed separately is the release of C-14 and Rn-222 from the repository in gaseous form, entering the biosphere via soil as a diffuse source. Pathways considered are doses from ingestion of C-14 and from inhalation of C-14 and Rn-222 outdoors as well as indoors. For these scenarios, specific dose calculations were carried out. The methods applied for these calculations and the

Review of SR 97

International Nuclear Information System (INIS)

Voss, C.I.

2000-01-01

generate fluxes, travel paths and travel times, it is dangerous to use unproven probability distributions as the basis for assessment of hydrogeologic impact on repository safety. There may be serious doubt that the fluxes and path values derived from SR 97 ground-water modeling are appropriate for determining near-field release and far-field transport because they are based on poorly founded probabilistic assumptions, on weak hydrogeologic structural models of the sites, and on static boundary conditions, despite the expectation of strong climate change effects. In this light, it is interesting that SR 97 directly used very little of the extensive ground-water modeling results, funneling all of the considerations and complexities for each site into a few selected values for use in release, transport, and dose calculations. It can be argued that, if these few values are all that are needed for performance assessment, they can be equally well determined by simple hydrogeologic scoping calculations for a site, rather than through the type of extensive effort applied in SR 97

Spent Fuel Dissolution and Source Term Modelling in Safety Assessment. Report from a Workshop. Synthesis and extended abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-05-15

This report describes a workshop that was organised by the Swedish Nuclear Power Inspectorate (SKI) for assessment of the handling of near-field radionuclide retention processes by the Swedish Nuclear Fuel and Waste Management Company (SKB). The general objective with this type of meeting is to improve the knowledge and awareness of recent developments and to provide preliminary review comments. A number of SKB reports provided the general background for the workshop discussions. One report addresses the release of radionuclides from spent fuel, another the concentration limits related to radionuclide solubility and a third buffer radionuclide sorption and migration parameters. These reports comprise a basis for the handling of the spent fuel, solubility and sorption processes in new complete safety assessment SR-Can. The discussion and analysis of these background reports at the workshop therefore provide an essential element of preparation for the planned review of SR-Can. The review comments provided in this report are nonetheless of a preliminary character since the SR-Can report was not available at the time of the workshop and details about the incorporation of various potential safety features into the entirety of safety assessment were not known. The present report sets out the detailed objectives and format of the workshop in Section 2. Section 3 provides a high-level overview of processes that need to be taken into account. In Section 4, there is a brief discussion about the chemical and physical environment near the engineered barriers. Section 5 gives a more detailed description of spent fuel processes that affect the radionuclide releases. In Section 6, the key issues for radionuclide chemistry and the estimation of concentration limits for various radionuclides are discussed. Section 7 discusses radionuclide sorption and migration in the buffer and Section 8 presents overall conclusions from the workshop.

Landscape dose conversion factors used in the safety assessment SR-Site

International Nuclear Information System (INIS)

Avila, Rodolfo; Ekstroem, Per-Anders; Aastrand, Per-Gustav

2010-12-01

In this report two types of Dose Conversion Factors have been derived: i) a Landscape Dose Conversion Factor (LDF) that is applicable to continuous long-term releases to the biosphere at a constant rate, and ii) a Landscape Dose Conversion Factor for pulse releases (LDF pulse) that is applicable to a radionuclide release that reaches the biosphere in a pulse within years to hundreds of years. In SR-Site these Dose Factors are multiplied with modelled release rates or pulse releases from the geosphere to obtain dose estimates used in assessment of compliance with the regulatory risk criterion. The LDFs were calculated for three different periods of the reference glacial cycle; a period of submerged conditions following the deglaciation, the temperate period, and a prolonged period of periglacial conditions. Additionally, LDFs were calculated for the global warming climate case. The LDF pulse was calculated only for temperate climate conditions. The LDF and LDF pulse can be considered as Best Estimate values, which can be used in calculations of Best Estimate values of doses to a representative individual of the most exposed group from potential releases from a future repository. A systematic analysis of the effects of system, model and parameter uncertainties on the LDFs has been carried out. This analysis has shown that the use of the derived LDF would lead to cautious or realistic dose estimates. The models and methods that were used for derivation of the LDFs and LDF pulse are also described in this report

Landscape dose conversion factors used in the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Avila, Rodolfo; Ekstroem, Per-Anders; Aastrand, Per-Gustav (Facilia AB (Sweden))

2010-12-15

In this report two types of Dose Conversion Factors have been derived: i) a Landscape Dose Conversion Factor (LDF) that is applicable to continuous long-term releases to the biosphere at a constant rate, and ii) a Landscape Dose Conversion Factor for pulse releases (LDF pulse) that is applicable to a radionuclide release that reaches the biosphere in a pulse within years to hundreds of years. In SR-Site these Dose Factors are multiplied with modelled release rates or pulse releases from the geosphere to obtain dose estimates used in assessment of compliance with the regulatory risk criterion. The LDFs were calculated for three different periods of the reference glacial cycle; a period of submerged conditions following the deglaciation, the temperate period, and a prolonged period of periglacial conditions. Additionally, LDFs were calculated for the global warming climate case. The LDF pulse was calculated only for temperate climate conditions. The LDF and LDF pulse can be considered as Best Estimate values, which can be used in calculations of Best Estimate values of doses to a representative individual of the most exposed group from potential releases from a future repository. A systematic analysis of the effects of system, model and parameter uncertainties on the LDFs has been carried out. This analysis has shown that the use of the derived LDF would lead to cautious or realistic dose estimates. The models and methods that were used for derivation of the LDFs and LDF pulse are also described in this report

Compliance demonstration: What can be reasonably expected from safety assessment for geological repositories?

International Nuclear Information System (INIS)

Zuidema, P.; Smith, P.; Sumerling, T.

1999-01-01

When licensing a nuclear facility, it is important to demonstrate that it will comply with regulatory limits (e.g. individual dose limits) and also show that sufficient attention has been paid to optimisation of facility design and operation, such that any associated radiological impacts will be as low as reasonably achievable (ALARA). In general, in demonstrating compliance, experience can be drawn from the performance of existing and similar facilities, and monitoring plans can be specified that will confirm that actual radiological discharges during operations are within authorised limits for the facility. This is also true in respect of the operational period of a geological repository. For the post-closure phase of a repository, however, it is also necessary to show that possible releases will remain acceptably low even at long times in the future when, it is assumed, control of the facility has lapsed and there is no method of either monitoring releases or taking remedial action in the case of unexpected events or releases. In addition, within each country, a deep geological repository will be a first-of-a-kind development so that compliance arguments can be expected to be rigorously tested without any assistance from the precedent of licensing of similar facilities nationally. This puts heavy, and quite unusual, burdens on the long-term safety assessment for a geological repository to develop a case that is sufficiently strong to demonstrate compliance. This paper focuses on the problem of demonstrating compliance with long-term safety requirements for a geological repository, and explores: the overall aims and special difficulties of demonstrating compliance for a geological repository; the role of safety assessment in demonstrating compliance; the scope for optimisation of a geological repository and importance of robustness and lessons learnt from the application of safety assessment. In addition, some issues requiring further discussion and clarification

International Expert Review of SRCan: Site Investigation Aspects. External review contribution in support of SKI's and SSI's review of SR-Can. INSITE/OVERSITE

Energy Technology Data Exchange (ETDEWEB)

2008-03-15

As a first evaluation of long-term safety for KBS-3 repositories at Forsmark and Laxemar, the SIG (Site Investigation Group) found SR-Can to be a well-produced and generally well-argued safety assessment. Overall, SKB is to be complimented on this project. Members of of the two groups INSITE and OVERSITE within the SIG had somewhat differing views on how well SKB had made use of the site data available at the end of the SDM 1.2 stage of investigations. This difference is less to do with the extent of site characterisation than of its use and application, reflecting the different levels of maturity of SKB's geosphere and biosphere assessment programmes. The more recent and current work on the sites means that our concerns expressed in this review should, to a large extent, be addressable in or prior to SR-Site, provided SKB is so minded. However, we acknowledge that some of the issues we raise will not be fully resolved until underground rock characterisation from excavations or longer records of surface conditions are available. There are also some key aspects of SKB's methodology still under development that would benefit from review prior to their use in SR-Site. More space in the currently pressing schedule would allow for this review and a consequent increase in confidence. In any case, the authorities should be aware that SKB may face residual programmatic risks, associated principally with the underground design and layout (and their knockon effects into performance), even after SR-Site. An early understanding of some of these relationships would be helped by a plan (at least on an outline level) of the underground characterisation programme. We also note that many engineering matters are still to be confronted, not least the EBS design and its implementation, along with the treatment of high stresses, if Forsmark is selected. However, our views on the nature of the SR-Can analysis and the way in which site data have been utilised in it (our

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

International Nuclear Information System (INIS)

1999-11-01

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

Thermodynamic assessment of the Sn–Sr system supported by first-principles calculations

International Nuclear Information System (INIS)

Zhao, Jingrui; Du, Yong; Zhang, Lijun; Wang, Aijun; Zhou, Liangcai; Zhao, Dongdong; Liang, Jianlie

2012-01-01

Highlights: ► All the literature data of Sn–Sr system is critically reviewed. ► First-principles calculation of enthalpy of formation is carried out for each compound. ► Thermodynamic parameters for Sn–Sr system are obtained by CALPHAD method. ► A hybrid approach of CALPHAD and first-principles calculations is recommended. - Abstract: A hybrid approach of CALPHAD and first-principles calculations was employed to perform a thermodynamic modeling of the Sn–Sr system. The experimental phase diagram and thermodynamic data available in the literature were critically reviewed. The enthalpies of formation for the 6 stoichiometric compounds (i.e. Sr 2 Sn, Sr 5 Sn 3 , SrSn, Sr 3 Sn 5 , SrSn 3 and SrSn 4 ) at 0 K were computed by means of first-principles calculations. These data were used as the experimental values in the optimization module PARROT in the subsequent CALPHAD assessment to provide thermodynamic parameters with sound physical meaning. A set of self-consistent thermodynamic parameters was finally obtained by considering reliable literature data and the first-principles computed results. Comprehensive comparisons between the calculated and measured quantities indicate that all the reliable experimental information can be satisfactorily accounted for by the present thermodynamic description.

Human factors in safety assessment. Safety culture assessment

International Nuclear Information System (INIS)

Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

1996-01-01

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

SR 97 - Identification and structuring of process

International Nuclear Information System (INIS)

Pers, K.; Skagius, K.; Soedergren, S.; Wiborgh, M.; Hedin, A.; Moren, L.; Sellin, P.; Stroem, A.; Pusch, R.; Bruno, J.

1999-12-01

This report documents work conducted in recent years to identify processes and interactions of importance to the evaluation of long-term safety of a KBS 3 type deep repository for spent nuclear fuel. Previous, partly undocumented work regarding interaction matrices is described as well as the THMC diagrams that have been used in the safety assessment SR 97. The coupling between the two sources of information is documented in a database. In the same database, the interaction matrices are briefly documented, while the processes in the THMC diagrams are more thoroughly documented in a special so called Process Report, which forms an important supporting document for SR 97

SR 97 - Identification and structuring of process

Energy Technology Data Exchange (ETDEWEB)

Pers, K.; Skagius, K.; Soedergren, S.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden); Hedin, A.; Moren, L.; Sellin, P.; Stroem, A. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Pusch, R. [Geodevelopment AB, Lund (Sweden); Bruno, J. [QuantiSci SL, Barcelona (Spain)

1999-12-01

This report documents work conducted in recent years to identify processes and interactions of importance to the evaluation of long-term safety of a KBS 3 type deep repository for spent nuclear fuel. Previous, partly undocumented work regarding interaction matrices is described as well as the THMC diagrams that have been used in the safety assessment SR 97. The coupling between the two sources of information is documented in a database. In the same database, the interaction matrices are briefly documented, while the processes in the THMC diagrams are more thoroughly documented in a special so called Process Report, which forms an important supporting document for SR 97.

Review of SKB's Work on Coupled THM Processes Within SR-Can. External review contribution in support of SKI's and SSI's review of SR-Can

Energy Technology Data Exchange (ETDEWEB)

Rutqvist, Jonny; Chin-Fu Tsang (Lawrence Berkeley National Laboratory, Berkeley, CA (US))

2008-03-15

In this report, we scrutinize the work by the Swedish Nuclear Fuel and Waste Management Company (SKB) related to coupled thermal, hydrological and mechanical (THM) processes within the SR-Can project. SR-Can is SKB's preliminary assessment of long-term safety for a KBS-3 nuclear waste repository, and is a preparation stage for the SR-Site assessment, the report that will be used in SKB's application for a final repository. We scrutinize SKB's work related to THM processes through review and detailed analysis, using an independent modeling tool. The modeling tool is applied to analyze coupled THM processes at the two candidate sites, Forsmark and Laxemar, using data defined in SKB's site description models for respective sites. In this report, we first provide a brief overview of SKB's work related to analysis of the evolution of coupled THM processes as presented in SRCan, as well as supporting documents. In this overview we also identify issues and assumptions that we then analyze using our modeling tool. The overview and subsequent independent model analysis addresses issues related to near-field behavior, such as buffer resaturation and the evolution of the excavation-disturbed zone, as well as far-field behavior, such as stress induced changes in hydrologic properties. Based on the review and modeling conducted in this report, we conclude by identifying a number of areas of weaknesses, where we believe further work and clarifications are needed. Some of the most important ones are summarized below: 1) We found that SKB's calculation of peak temperature might not have been conducted for the most conservative case of extreme drying of the buffer under dry rock conditions and an unexpectedly high thermal diffusion coefficient. Our alternative analysis indicates that temperatures close to 100 might be achieved under unfavorable (and perhaps unexpected) conditions in which the buffer is dried to below 20% near the canister. We believe

Long-term safety for the final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project

Energy Technology Data Exchange (ETDEWEB)

2011-03-15

The central conclusion of the safety assessment SR-Site is that a KBS-3 repository that fulfils long-term safety requirements can be built at the Forsmark site. This conclusion is reached because the favourable properties of the Forsmark site ensure the required long-term durability of the barriers of the KBS-3 repository. In particular, the copper canisters with their cast iron inserts have been demonstrated to provide a sufficient resistance to the mechanical and chemical loads to which they may be subjected in the repository environment. The conclusion is underpinned by: - The reliance of the KBS-3 repository on i) a geological environment that exhibits long-term stability with respect to properties of importance for long-term safety, i.e. mechanical stability, low groundwater flow rates at repository depth and the absence of high concentrations of detrimental components in the groundwater, and ii) the choice of naturally occurring materials (copper and bentonite clay) for the engineered barriers that are sufficiently durable in the repository environment to provide the barrier longevity required for safety. - The understanding, through decades of research at SKB and in international collaboration, of the phenomena that affect long-term safety, resulting in a mature knowledge base for the safety assessment. - The understanding of the characteristics of the site through several years of surface-based investigations of the conditions at depth and of scientific interpretation of the data emerging from the investigations, resulting in a mature model of the site, adequate for use in the safety assessment. - The detailed specifications of the engineered parts of the repository and the demonstration of how components fulfilling the specifications are to be produced in a quality assured manner, thereby providing a quality assured initial state for the safety assessment. The detailed analyses demonstrate that canister failures in a one million year perspective are rare

Long-term safety for the final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project

International Nuclear Information System (INIS)

2011-03-01

The central conclusion of the safety assessment SR-Site is that a KBS-3 repository that fulfils long-term safety requirements can be built at the Forsmark site. This conclusion is reached because the favourable properties of the Forsmark site ensure the required long-term durability of the barriers of the KBS-3 repository. In particular, the copper canisters with their cast iron inserts have been demonstrated to provide a sufficient resistance to the mechanical and chemical loads to which they may be subjected in the repository environment. The conclusion is underpinned by: - The reliance of the KBS-3 repository on i) a geological environment that exhibits long-term stability with respect to properties of importance for long-term safety, i.e. mechanical stability, low groundwater flow rates at repository depth and the absence of high concentrations of detrimental components in the groundwater, and ii) the choice of naturally occurring materials (copper and bentonite clay) for the engineered barriers that are sufficiently durable in the repository environment to provide the barrier longevity required for safety. - The understanding, through decades of research at SKB and in international collaboration, of the phenomena that affect long-term safety, resulting in a mature knowledge base for the safety assessment. - The understanding of the characteristics of the site through several years of surface-based investigations of the conditions at depth and of scientific interpretation of the data emerging from the investigations, resulting in a mature model of the site, adequate for use in the safety assessment. - The detailed specifications of the engineered parts of the repository and the demonstration of how components fulfilling the specifications are to be produced in a quality assured manner, thereby providing a quality assured initial state for the safety assessment. The detailed analyses demonstrate that canister failures in a one million year perspective are rare

Report on nuclear energy in SR Slovenia; Porocilo o uporabi jedrske energije v SR Sloveniji

Energy Technology Data Exchange (ETDEWEB)

NONE

1987-07-01

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1987.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-11-01

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

Review of SKB's reporting of SR 97

International Nuclear Information System (INIS)

Tiren, S.A.

2000-01-01

The task of the safety assessment is to show 'that the repository has been designed with sufficient margins to be safe in spite of the incomplete knowledge available'. SKB mentions 'confidence in the results' as an important aspect. For the layman, confidence in the information presented is of the greatest importance. This means that the presentation of the safety assessment should be of a high visual standard with respect to descriptions of processes and events. SKB presented the purpose of SR 97 as four points. From a geological-structural geological viewpoint, the following can be mentioned: 1. Methodology for evaluating the geometry of the structural patterns of the bedrock at a depth of 500 m exists in general, but there are also certain deficiencies. Examples include determining the position of the individual structures and obtaining information. The importance of the bedrock is not completely clear. What is related to the rock type and what is related to the geological environment (including the geological evolution)? A clearer and more systematic compilation of data used in the safety assessment is required. 2. Work involving alternative models and evaluations of how well the models explain the collected data would be appropriate. The geological and structural models for a site are included as the base data in calculations and other modelling conducted in connection with the characterization of a site. 3. SR 97 does not provide any detailed information on site investigations. 4. The function of the bedrock as an external barrier is shown. However, to a certain extent, information on how this barrier can be affected by aseismic creep movements along fracture structures and the impact of selective erosion along such structures is lacking. Furthermore, criteria for properties of the rock volumes where deposition drifts are planned should be expanded, such as by determining a minimum width and suitable length-width relations

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

scope of the quantitative safety assessment. These arguments include: Support from natural and anthropogenic analogues for both key process understanding and total system performance. Comparison of the methodology and results with the earlier TILA-99 and SR-Can safety assessments, as well as other international safety assessments, to ensure completeness, consistency and reasonableness of the present assessment. Use of safety indicators other than dose and activity to avoid uncertainties in future human lifestyles and also in geological processes on very long timescales. Consideration of the calculation results from a wider perspective to consider significance of their impact compared to other risks

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Prediction of concentration and model validation - key issues in assessment of long term safety for radioactive waste disposal

International Nuclear Information System (INIS)

Xu, S.; Dverstorp, B.; Woerman, A.

2008-01-01

Post-closure safety assessments for nuclear waste repositories involve radioecological modelling for en,underground source term. In this paper we discuss critical aspects concerning process understanding and justification of simplified radioecological models used for such safety assessments. This study is part of the Swedish Radiation Protection Authority's (SSI) work on reviewing the Swedish Nuclear Fuel and Waste Management Co's (SKB) most recent safety assessment, SR-Can. One of the most challenging tasks in assessments of environmental doses and risk from an underground repository is to estimate radionuclide activity concentrations in various geologic strata in the future. For example, little is known about transport pathways through the quaternary deposits to the discharge points in surface waters and other recipients in the biosphere. Traditionally simplified compartmental models are used in safety assessment to describe the fate of radio-nuclides in surface environment. The possibility to test such models against more detailed process models and site specific data is of key importance for confidence in the safety assessment. As part of SSI's review of SR-Can, alternative modelling approaches were developed to explore the importance of transport process descriptions in the assessment models. The modelling results were compared with the Landscape Dose Factors (LDFs) derived by SKB in SR-Can. LDFs is a new methodology adapted by SKB in SR-Can. The LDFs are defined in the units of Sv/y per Bq/y and express all the radiological information about individual epository sites and ecosystems as a single, radionuclide-specific, number that relates geosphere releases to radiological dose. Further, we suggest a method for validating model parameters using data from field tracer tests. In two companion papers we present the underlying model framework for pathway analyses and a newly developed numerical module within the numerical software Ecolego Toolbox. Transport models

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

that lie outside the scope of the quantitative safety assessment. These arguments include: Support from natural and anthropogenic analogues for both key process understanding and total system performance. Comparison of the methodology and results with the earlier TILA-99 and SR-Can safety assessments, as well as other international safety assessments, to ensure completeness, consistency and reasonableness of the present assessment. Use of safety indicators other than dose and activity to avoid uncertainties in future human lifestyles and also in geological processes on very long timescales. Consideration of the calculation results from a wider perspective to consider significance of their impact compared to other risks

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

Report on nuclear energy in SR Slovenia

International Nuclear Information System (INIS)

1987-01-01

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1987.

Chemical conditions in present and future ecosystems in Forsmark - implications for selected radionuclides in the safety assessment SR-Site

International Nuclear Information System (INIS)

Troejbom, Mats; Grolander, Sara

2010-12-01

This report is a background report for the biosphere analysis of the SR-Site Safety Assessment. This work aims to describe the future development of the chemical conditions at Forsmark, based on the present chemical conditions at landscape level taking landscape development and climate cases into consideration. The results presented contribute to the overall understanding of the present and future chemistry in the Forsmark area, and specifically, to the understanding of the behaviour of some selected radionuclides in the surface system. The future development of the chemistry at the site is qualitatively discussed with focus on the interglacial within the next 10,000 years. The effects on the chemical environment of future climate cases as Global Warming and cold permafrost climates are also briefly discussed. The work is presented in two independent parts describing background radionuclide activities in the Forsmark area and the distribution and behaviour of a large number of stable elements in the landscape. In a concluding section, implications of the future chemical environment of a selection of radionuclides important in the Safety Assessment are discussed based on the knowledge of stable elements. The broad range of elements studied show that there are general and expected patterns for the distribution and behaviour in the landscape of different groups of elements. Mass balances reveal major sources and sinks, pool estimations show where elements are accumulated in the landscape and estimations of time-scales give indications of the potential future development. This general knowledge is transferred to radionuclides not measured in order to estimate their behaviour and distribution in the landscape. It could be concluded that the future development of the chemical environment in the Forsmark area might affect element specific parameters used in de radionuclide model in different directions depending on element. The alternative climate cases, Global Warming

Chemical conditions in present and future ecosystems in Forsmark - implications for selected radionuclides in the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Troejbom, Mats (Mats Troejbom Konsult AB (Sweden)); Grolander, Sara (Facilia AB (Sweden))

2010-12-15

This report is a background report for the biosphere analysis of the SR-Site Safety Assessment. This work aims to describe the future development of the chemical conditions at Forsmark, based on the present chemical conditions at landscape level taking landscape development and climate cases into consideration. The results presented contribute to the overall understanding of the present and future chemistry in the Forsmark area, and specifically, to the understanding of the behaviour of some selected radionuclides in the surface system. The future development of the chemistry at the site is qualitatively discussed with focus on the interglacial within the next 10,000 years. The effects on the chemical environment of future climate cases as Global Warming and cold permafrost climates are also briefly discussed. The work is presented in two independent parts describing background radionuclide activities in the Forsmark area and the distribution and behaviour of a large number of stable elements in the landscape. In a concluding section, implications of the future chemical environment of a selection of radionuclides important in the Safety Assessment are discussed based on the knowledge of stable elements. The broad range of elements studied show that there are general and expected patterns for the distribution and behaviour in the landscape of different groups of elements. Mass balances reveal major sources and sinks, pool estimations show where elements are accumulated in the landscape and estimations of time-scales give indications of the potential future development. This general knowledge is transferred to radionuclides not measured in order to estimate their behaviour and distribution in the landscape. It could be concluded that the future development of the chemical environment in the Forsmark area might affect element specific parameters used in de radionuclide model in different directions depending on element. The alternative climate cases, Global Warming

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

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)

Food and drinking water safety: Can risk assessment help us to get our priorities right?

International Nuclear Information System (INIS)

Denner, W.H.B.

1992-01-01

Huge resources are devoted worldwide by governments and food producers to ensure that food and water are produced with due regard to the safety of consumers. This inevitably involves some form of risk assessment but in the field of food safety a formalised process of risk assessment has been slow to develop. An ad hoc mosaic or approaches has evolved which varies not only between countries but sometimes within countries as well. This may not be unexpected considering the vast variety of kinds of food hazards (table 1). Not only do food-related hazards themselves vary widely, so do the effects which they can cause. For example microorganisms can cause mild stomach upsets or death within a few hours depending upon the organism involved. For chemical contaminants in food the potential effects are usually less acute although no less serious. Many of the chemicals of concern are believed to be carcinogens whose effects might only be realised after many years of exposure. Nutritional imbalances may result in an increased risk from diseases, such as heart disease and cancer, which can also arise from other causes. In these latter examples it is often difficult to relate cause to effect even when extensive epidemiological evidence is available. It is important to understand the enormous diversity in possible food-related hazards before describing the assessment of risks associated with them. This great diversity makes it unlikely that any single approach to risk assessment can suit all situations. This means that making comparisons between risks from different hazards is extremely difficult. In fact trying to allocate resources in a logical way between all the different kinds of food-related hazards is a major problem in itself. For with finite resources there is always the danger of finding that focusing on one area of concern results in a potential risk elsewhere being neglected. The aim of this paper is to take a general look at some of the issues facing those with

Food and drinking water safety: Can risk assessment help us to get our priorities right?

Energy Technology Data Exchange (ETDEWEB)

Denner, W H.B. [Ministry of Agriculture, Fisheries and Food, London (United Kingdom)

1992-07-01

Huge resources are devoted worldwide by governments and food producers to ensure that food and water are produced with due regard to the safety of consumers. This inevitably involves some form of risk assessment but in the field of food safety a formalised process of risk assessment has been slow to develop. An ad hoc mosaic or approaches has evolved which varies not only between countries but sometimes within countries as well. This may not be unexpected considering the vast variety of kinds of food hazards (table 1). Not only do food-related hazards themselves vary widely, so do the effects which they can cause. For example microorganisms can cause mild stomach upsets or death within a few hours depending upon the organism involved. For chemical contaminants in food the potential effects are usually less acute although no less serious. Many of the chemicals of concern are believed to be carcinogens whose effects might only be realised after many years of exposure. Nutritional imbalances may result in an increased risk from diseases, such as heart disease and cancer, which can also arise from other causes. In these latter examples it is often difficult to relate cause to effect even when extensive epidemiological evidence is available. It is important to understand the enormous diversity in possible food-related hazards before describing the assessment of risks associated with them. This great diversity makes it unlikely that any single approach to risk assessment can suit all situations. This means that making comparisons between risks from different hazards is extremely difficult. In fact trying to allocate resources in a logical way between all the different kinds of food-related hazards is a major problem in itself. For with finite resources there is always the danger of finding that focusing on one area of concern results in a potential risk elsewhere being neglected. The aim of this paper is to take a general look at some of the issues facing those with

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)

Thermodynamic assessment of the Pb-Sr system

Directory of Open Access Journals (Sweden)

Zhang H.

2017-01-01

Full Text Available The Pb-Sr system has been critically reviewed and modeled by means of the CALPHAD (CALculation of PHAse Diagrams approach. It contains seven stoichiometric compounds, i.e. SrPb3, Sr3Pb5, Sr2Pb3, SrPb, Sr5Pb4, Sr5Pb3 and Sr2Pb, in which the SrPb3 and Sr2Pb phases melt congruently, and the other five phases form via peritectic reactions. The enthalpies of formation for the intermetallic compounds at 0 K are provided by first-principles calculations. The liquid, fcc and bcc phases are modeled as substitutional solution phases. Both Redlich-Kister and exponential polynomials are used to describe the excess Gibbs energy of the liquid. Two sets of self-consistent thermodynamic parameters are obtained by considering reliable experimental data and the computed enthalpies of formation. Comprehensive comparisons between the calculated and measured phase diagram and thermodynamic data show that the experimental information is satisfactorily accounted for by the present thermodynamic description.

Spent Fuel Dissolution and Source Term Modelling in Safety Assessment. Report from a Workshop at Sigtunahoejden Hotel and Conference, Sigtuna, Sweden May 17-19, 2006. Synthesis and extended abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-05-15

This report describes a workshop that was organised by the Swedish Nuclear Power Inspectorate (SKI) for assessment of the handling of near-field radionuclide retention processes by the Swedish Nuclear Fuel and Waste Management Company (SKB). The general objective with this type of meeting is to improve the knowledge and awareness of recent developments and to provide preliminary review comments. A number of SKB reports provided the general background for the workshop discussions. One report addresses the release of radionuclides from spent fuel, another the concentration limits related to radionuclide solubility and a third buffer radionuclide sorption and migration parameters. These reports comprise a basis for the handling of the spent fuel, solubility and sorption processes in new complete safety assessment SR-Can. The discussion and analysis of these background reports at the workshop therefore provide an essential element of preparation for the planned review of SR-Can. The review comments provided in this report are nonetheless of a preliminary character since the SR-Can report was not available at the time of the workshop and details about the incorporation of various potential safety features into the entirety of safety assessment were not known. The present report sets out the detailed objectives and format of the workshop in Section 2. Section 3 provides a high-level overview of processes that need to be taken into account. In Section 4, there is a brief discussion about the chemical and physical environment near the engineered barriers. Section 5 gives a more detailed description of spent fuel processes that affect the radionuclide releases. In Section 6, the key issues for radionuclide chemistry and the estimation of concentration limits for various radionuclides are discussed. Section 7 discusses radionuclide sorption and migration in the buffer and Section 8 presents overall conclusions from the workshop.

Spent Fuel Dissolution and Source Term Modelling in Safety Assessment. Report from a Workshop at Sigtunahoejden Hotel and Conference, Sigtuna, Sweden May 17-19, 2006. Synthesis and extended abstracts

International Nuclear Information System (INIS)

2007-05-01

This report describes a workshop that was organised by the Swedish Nuclear Power Inspectorate (SKI) for assessment of the handling of near-field radionuclide retention processes by the Swedish Nuclear Fuel and Waste Management Company (SKB). The general objective with this type of meeting is to improve the knowledge and awareness of recent developments and to provide preliminary review comments. A number of SKB reports provided the general background for the workshop discussions. One report addresses the release of radionuclides from spent fuel, another the concentration limits related to radionuclide solubility and a third buffer radionuclide sorption and migration parameters. These reports comprise a basis for the handling of the spent fuel, solubility and sorption processes in new complete safety assessment SR-Can. The discussion and analysis of these background reports at the workshop therefore provide an essential element of preparation for the planned review of SR-Can. The review comments provided in this report are nonetheless of a preliminary character since the SR-Can report was not available at the time of the workshop and details about the incorporation of various potential safety features into the entirety of safety assessment were not known. The present report sets out the detailed objectives and format of the workshop in Section 2. Section 3 provides a high-level overview of processes that need to be taken into account. In Section 4, there is a brief discussion about the chemical and physical environment near the engineered barriers. Section 5 gives a more detailed description of spent fuel processes that affect the radionuclide releases. In Section 6, the key issues for radionuclide chemistry and the estimation of concentration limits for various radionuclides are discussed. Section 7 discusses radionuclide sorption and migration in the buffer and Section 8 presents overall conclusions from the workshop

New Safety rules

CERN Multimedia

Safety Commission

2008-01-01

The revision of CERN Safety rules is in progress and the following new Safety rules have been issued on 15-04-2008: Safety Procedure SP-R1 Establishing, Updating and Publishing CERN Safety rules: http://cern.ch/safety-rules/SP-R1.htm; Safety Regulation SR-S Smoking at CERN: http://cern.ch/safety-rules/SR-S.htm; Safety Regulation SR-M Mechanical Equipment: http://cern.ch/safety-rules/SR-M.htm; General Safety Instruction GSI-M1 Standard Lifting Equipment: http://cern.ch/safety-rules/GSI-M1.htm; General Safety Instruction GSI-M2 Standard Pressure Equipment: http://cern.ch/safety-rules/GSI-M2.htm; General Safety Instruction GSI-M3 Special Mechanical Equipment: http://cern.ch/safety-rules/GSI-M3.htm. These documents apply to all persons under the Director General’s authority. All Safety rules are available at the web page: http://www.cern.ch/safety-rules The Safety Commission

Sequential separation of cs, ca and ba for 90sr assessment

International Nuclear Information System (INIS)

Dianu, M.; Bucur, C.

2015-01-01

A two-steps chemical treatment technique for strontium assessment from aqueous samples is described in this paper. The method was applied to simulated samples containing stable elements of Ni, Cs, Ca, Ba, Mn, Fe, Co and Eu. The transition elements (Ni, Mn, Fe, Co, Eu) were precipitated as hydroxides, followed by alkaline-earth metals separation (Ca, Ba) as carbonates. Finally, the Sr was purified by extraction chromatography using Triskem International Sr resin. The strength of Sr sorption in nitric acid increases with increasing acid concentration, and the optimal bonding strength is achieved in 8 M HNO3. The combination of successive precipitations with extraction chromatography for complete removal of other interferences from Sr matrix leads to good recovery and decontamination factor values. (authors)

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)

SR'97 peer review: feedback on the integration of comments in the stepwise process

International Nuclear Information System (INIS)

Hedin, Allan

2002-01-01

Allan Hedin (SKB, Sweden) described recent reviews of SR'97 by a NEA review team and by the Swedish regulatory authorities, SKI and SKB. The NEA review team concluded that KBS-3 was a sound disposal concept and that SR 97 provided a sensible illustration of the potential safety of that concept. On that basis SKB's desire to move to a site selection phase was well founded. It was suggested that there should in future be more frequent, iterative, safety assessments and it would be preferable if more formal scenario selection techniques were used. The regulatory authorities also found that the KBS-3 concept was soundly based but indicated that an exhaustive analysis of the concept would be required in due course taking account of information from site investigations and testing of the proposed engineered barriers. They were generally satisfied with the assessment methodologies described in SR-97, though some deficiencies were identified with regard to scenario development, the methodology for probabilistic calculations, methods for feedback to site investigation and design and the programme for future assessments

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

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.

OSART Independent Safety Culture Assessment (ISCA) Guidelines

International Nuclear Information System (INIS)

2016-01-01

in this publication follows the same principles as the IAEA methodology for safety culture self-assessments, but has one more essential data collection source, as it includes the OSART team’s data findings in the analysis. This publication can also be used whenever independent safety culture assessments are performed as a standalone or as add-on modules for other types of safety review service. Nevertheless, an integrated approach helps to ensure diversity of competences, and so the assessment addresses all aspects of nuclear safety. This publication updates IAEA Services Series No. 16, SCART Guidelines

Evaluation of SKB's report 'Deep repository for spent nuclear fuel: SR 97 - Post-closure safety', Focusing on the assessment of transport processes in the geosphere

International Nuclear Information System (INIS)

Woerman, A.; Shulan Xu

2000-01-01

This report describes a critical review of the safety assessment performed on the final repository for nuclear waste in Sweden that is proposed by SKB in 'Deep Repository for Spent Nuclear Fuel: SR 97 - Post-closure Safety'. The review was requested by the Swedish Nuclear Power Inspectorate (SKI). The waste repository consists of several barriers that work together with the purpose of delaying radionuclide migration and reducing the activity that eventually affects the biosphere. A main criticism is the lack of a formal risk analysis and uncertainties in several analyses that make it difficult to comprehend the overall risk of the repository. A formal risk analysis should comprise a probabilistic treatment of all components included in the system. This is not the case in the SKB's report since the probabilistic analyses are limited only to certain aspects. The use of conservative model parameters are not a substitute for risk analysis nor can they compensate for possible model biases. Bias can be expected in most of the existing models of radionuclide migration in fractured bedrock. SKB should present a clear comparison on the importance of the different barrier components (uranium-dioxide matrix, copper canister, buffer and bedrock) on the retardation of radionuclides. It is unclear as to what extent the capacity of the bedrock to retain migrating radionuclides is critical to the capacity of the repository. A large part of the SR 97 report is focused on retardation processes in bedrock and a reader can interpret this as the technical weight given on retardation in the bedrock. However, with the present state of knowledge, it is our opinion that we cannot with an acceptable degree of accuracy predict the radionuclide transport in bedrock or quantify risk levels associated with radioactivity in the biosphere. There are large uncertainties concerning the way by which sorption processes should be formulated and the impact of colloids on the transport that can be

Developing design premises for a KBS-3V repository based on results from the safety assessment - 16027

International Nuclear Information System (INIS)

Andersson, Johan; Hedin, Allan

2009-01-01

As a part of the planned license application for a final repository for spent nuclear fuel the Swedish Nuclear Fuel and Waste Management Co. (SKB), has developed design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel. The purpose is to provide requirements from a long term safety aspect, to form the basis for the development of the reference design of the repository and to justify that design. Design premises typically concern specification on what mechanical loads the barriers must withstand, restrictions on the composition of barrier materials or acceptance criteria for the various underground excavations. These design constraints, if all fulfilled by the actual design, should form a good basis for demonstrating repository safety. The justification for these design premises is derived from SKB's most recent safety assessment SR-Can complemented by a few additional analyses. Some of the design premises may be modified in future stages of SKB's program, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. (authors)

Experiment data report for semiscale Mod-2A primary feed and bleed experiment series (Tests S-SR-1 and S-SR-2)

International Nuclear Information System (INIS)

Fogdall, S.P.

1982-10-01

This report presents test data recorded for Tests S-SR-1 and S-SR-2 of the Semiscale Mod-2A Primary Feed and Bleed Tests. These tests are part of a series of Semiscale tests that investigate the thermal-hydraulic phenomena resulting from a hypothesized loss-of-coolant accident (LOCA) or abnormal operating transient. These tests provide experimental data for assessing the analytical capability of computer codes used in LOCA and operational transient analysis. The primary objectives of Tests S-SR-1 and -2 were to provide data on primary system recovery through the use of primary feed and bleed cooling, with no heat transfer to the secondaries. Data was obtained using high- and low-head pump curves for the safety injection (SI) pumps. This report presents the uninterpreted data from Tests S-SR-1 and -2 for analysis. The data, presented as graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent

Assessment of Safety Parameters for Radiological Explosion Based on Gaussian Dispersion Model

Energy Technology Data Exchange (ETDEWEB)

Pandey, Alok [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Yu, Hyungjoon; Kim, Hong Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

These sources if used with explosive (called RDD - radiological dispersion device), can cause dispersion of radioactive material resulting in public exposure and contamination of the environment. Radiological explosion devices are not weapons for the mass destruction like atom bombs, but can cause the death of few persons and contamination of large areas. The reduction of the threat of radiological weapon attack by terrorist groups causing dispersion of radioactive material is one of the priority tasks of the IAEA Nuclear Safety and Security Program.Emergency preparedness is an essential part for reducing and mitigating radiological weapon threat. Preliminary assessment of dispersion study followed by radiological explosion and its quantitative effect will be helpful for the emergency preparedness team for an early response. The effect of the radiological dispersion depends on various factors like radioisotope, its activity, physical form, amount of explosive used and meteorological factors at the time of an explosion. This study aim to determine the area affected by the radiological explosion as pre assessment to provide feedback to emergency management teams for handling and mitigation the situation after an explosion. Most practical scenarios of radiological explosion are considered with conservative approach for the assessment of the area under a threat for emergency handling and management purpose. Radioisotopes under weak security controls can be used for a radiological explosion to create terror and socioeconomic threat for the public. Prior assessment of radiological threats is helpful for emergency management teams to take prompt decision about evacuation of the affected area and other emergency handling actions. Comparable activities of Co-60 source used in radiotherapy and Sr-90 source of disused and orphaned RTGs with two different quantities of TNT were used for the scenario development of radiological explosion. In the Basic Safety Standard (BSS

Assessment of Safety Parameters for Radiological Explosion Based on Gaussian Dispersion Model

International Nuclear Information System (INIS)

Pandey, Alok; Yu, Hyungjoon; Kim, Hong Suk

2014-01-01

These sources if used with explosive (called RDD - radiological dispersion device), can cause dispersion of radioactive material resulting in public exposure and contamination of the environment. Radiological explosion devices are not weapons for the mass destruction like atom bombs, but can cause the death of few persons and contamination of large areas. The reduction of the threat of radiological weapon attack by terrorist groups causing dispersion of radioactive material is one of the priority tasks of the IAEA Nuclear Safety and Security Program.Emergency preparedness is an essential part for reducing and mitigating radiological weapon threat. Preliminary assessment of dispersion study followed by radiological explosion and its quantitative effect will be helpful for the emergency preparedness team for an early response. The effect of the radiological dispersion depends on various factors like radioisotope, its activity, physical form, amount of explosive used and meteorological factors at the time of an explosion. This study aim to determine the area affected by the radiological explosion as pre assessment to provide feedback to emergency management teams for handling and mitigation the situation after an explosion. Most practical scenarios of radiological explosion are considered with conservative approach for the assessment of the area under a threat for emergency handling and management purpose. Radioisotopes under weak security controls can be used for a radiological explosion to create terror and socioeconomic threat for the public. Prior assessment of radiological threats is helpful for emergency management teams to take prompt decision about evacuation of the affected area and other emergency handling actions. Comparable activities of Co-60 source used in radiotherapy and Sr-90 source of disused and orphaned RTGs with two different quantities of TNT were used for the scenario development of radiological explosion. In the Basic Safety Standard (BSS

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

SR 97: Post-closure safety for a KBS 3 type deep repository for spent nuclear fuel

International Nuclear Information System (INIS)

Hedin, A.; Kautsky, U.

2000-03-01

Prior to coming site investigations for siting of a deep repository for spent nuclear fuel, SKB has carried out the long-term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion. The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. (author)

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

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)

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

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

Review of SR-Can: Evaluation of SKB's handling of spent fuel performance, radionuclide chemistry and geosphere transport parameters. External review contribution in support of SKI's and SSI's review of SR-Can

International Nuclear Information System (INIS)

Stenhouse, Mike; Jegou, Christophe; Brown, Paul; Meinrath, Guenther; Nitsche, Heino; Ekberg, Christian

2008-03-01

SR-Can covers the containment phase of the KBS-3 barriers as well as the consequences of releases of radionuclides to the rock and eventually the biosphere (after complete containment within fuel canisters has partially failed). The aim of this report is to provide a range of review comments with respect to those parameters related to spent fuel performance as well as radionuclide chemistry and transport. These parameter values are used in the quantification of consequences due to release of radionuclides from potentially leaking canisters. The report does not cover modelling approaches used for quantification of consequences. However, modelling used to derive parameter values is to some extent addressed (such as calculation of maximum radionuclide concentration due to formation of solubility limiting phases). The following are the key highlights and comments generated in the course of the review: Inconsistencies exist between recommendations provided in technical reports and those quoted in the Data Report. One of the reasons for such inconsistencies has been the timing of different pieces of research. It is hoped that the timing of contributions to SR-Site will be such that these inconsistencies can be avoided. Sensitivity analyses need to be carried out and reported in a number of areas to support some of the assumptions or decisions made in the assessment calculations. The likelihood is that SKB has performed many of the sensitivity analyses identified in different parts of this report, but these need to be reported, preferably to complement the recommendations provided

Safety Assessment for Decommissioning

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

In the past few decades, international guidance has been developed on methods for assessing the safety of predisposal and disposal facilities for radioactive waste. More recently, it has been recognized that there is also a need for specific guidance on safety assessment in the context of decommissioning nuclear facilities. The importance of safety during decommissioning was highlighted at the International Conference on Safe Decommissioning for Nuclear Activities held in Berlin in 2002 and at the First Review Meeting of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management in 2003. At its June 2004 meeting, the Board of Governors of the IAEA approved the International Action Plan on Decommissioning of Nuclear Facilities (GOV/2004/40), which called on the IAEA to: ''establish a forum for the sharing and exchange of national information and experience on the application of safety assessment in the context of decommissioning and provide a means to convey this information to other interested parties, also drawing on the work of other international organizations in this area''. In response, in November 2004, the IAEA launched the international project Evaluation and Demonstration of Safety for Decommissioning of Facilities Using Radioactive Material (DeSa) with the following objectives: -To develop a harmonized approach to safety assessment and to define the elements of safety assessment for decommissioning, including the application of a graded approach; -To investigate the practical applicability of the methodology and performance of safety assessments for the decommissioning of various types of facility through a selected number of test cases; -To investigate approaches for the review of safety assessments for decommissioning activities and the development of a regulatory approach for reviewing safety assessments for decommissioning activities and as a basis for regulatory decision making; -To provide a forum

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 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

Comment on the internal consistency of thermodynamic databases supporting repository safety assessments

International Nuclear Information System (INIS)

Arthur, R.C.

2001-11-01

This report addresses the concept of internal consistency and its relevance to the reliability of thermodynamic databases used in repository safety assessments. In addition to being internally consistent, a reliable database should be accurate over a range of relevant temperatures and pressures, complete in the sense that all important aqueous species, gases and solid phases are represented, and traceable to original experimental results. No single definition of internal consistency need to be universally accepted as the most appropriate under all conditions, however. As a result, two databases that are each internally consistent may be inconsistent with respect to each other, and a database derived from two or more such databases must itself be internally inconsistent. The consequences of alternative definitions that are reasonably attributable to the concept of internal consistency can be illustrated with reference to the thermodynamic database supporting SKB's recent SR 97 safety assessment. This database is internally inconsistent because it includes equilibrium constants calculated over a range of temperatures: using conflicting reference values for some solids, gases and aqueous species that are common to two internally consistent databases (the OECD/NEA database for radioelements and SUPCRT databases for non-radioactive elements) that serve as source databases for the SR 97 TDB, using different definitions in these source databases of standard states for condensed phases and aqueous species, based on different mathematical expressions used in these source databases representing the temperature dependence of the heat capacity, and based on different chemical models adopted in these source databases for the aqueous phase. The importance of such inconsistencies must be considered in relation to the other database reliability criteria noted above, however. Thus, accepting a certain level of internal inconsistency in a database it is probably preferable to use a

Comment on the internal consistency of thermodynamic databases supporting repository safety assessments

Energy Technology Data Exchange (ETDEWEB)

Arthur, R.C. [Monitor Scientific, LLC, Denver, CO (United States)

2001-11-01

This report addresses the concept of internal consistency and its relevance to the reliability of thermodynamic databases used in repository safety assessments. In addition to being internally consistent, a reliable database should be accurate over a range of relevant temperatures and pressures, complete in the sense that all important aqueous species, gases and solid phases are represented, and traceable to original experimental results. No single definition of internal consistency need to be universally accepted as the most appropriate under all conditions, however. As a result, two databases that are each internally consistent may be inconsistent with respect to each other, and a database derived from two or more such databases must itself be internally inconsistent. The consequences of alternative definitions that are reasonably attributable to the concept of internal consistency can be illustrated with reference to the thermodynamic database supporting SKB's recent SR 97 safety assessment. This database is internally inconsistent because it includes equilibrium constants calculated over a range of temperatures: using conflicting reference values for some solids, gases and aqueous species that are common to two internally consistent databases (the OECD/NEA database for radioelements and SUPCRT databases for non-radioactive elements) that serve as source databases for the SR 97 TDB, using different definitions in these source databases of standard states for condensed phases and aqueous species, based on different mathematical expressions used in these source databases representing the temperature dependence of the heat capacity, and based on different chemical models adopted in these source databases for the aqueous phase. The importance of such inconsistencies must be considered in relation to the other database reliability criteria noted above, however. Thus, accepting a certain level of internal inconsistency in a database it is probably preferable to

AMBER and Ecolego Intercomparisons using Calculations from SR 97

International Nuclear Information System (INIS)

Maul, Philip; Robinson, Peter; Avila, Rodolfo; Broed, Robert; Pereira, Antonio

2003-08-01

. The residual differences in the other calculations are due to small differences in the interpretation of the SR 97 models and data. 3. The results of the deterministic calculations for the geosphere, where AMBER used a compartment model and Ecolego a one-dimensional contaminant transport model, were also generally in agreement to within one significant figure, except where radionuclides fluxes are very low (e.g., for 239 Pu), where the compartmental approximation in AMBER can overestimate the output flux. 4. In several instances the SR 97 documentation was not sufficiently clear to enable the implementation of the models in AMBER and Ecolego to be unambiguous. Examples include the structure of the coastal biosphere model and the algorithms used in some of the dose calculations. 5. Some key issues for the SR 97 assessment have been identified; these provide an indication of where detailed scrutiny will be required of any safety case made by SKB for a deep repository

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)

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

Expert Opinion in SR 97 and the SKI/SSI Joint Review of SR 97

Energy Technology Data Exchange (ETDEWEB)

Hora, Stephen

2002-09-01

The role of sensitivity and uncertainty analyses for radioactive waste disposal assessments is reviewed. The report covers a description of the these concepts were applied in the authorities' review of the safety report SR 97. With regard to the use of expert knowledge, the most significant weakness of SR 97 is absence of any standards, procedures, and even definitions for expert judgment. This situation needs to be dealt with by SKB in the near future as it denigrates the portions of the study that are well done. In developing expert judgment processes, SSI should ensure that SKB creates procedures that guarantee traceability and transparency. This will become very important as the repository system matures and receives greater public scrutiny. Both in the area of scenario creation and expert judgement, there are processes that have gained international acceptance. It would be in the best interest of SKB, and the public, to adhere these accepted approaches.

Expert Opinion in SR 97 and the SKI/SSI Joint Review of SR 97

International Nuclear Information System (INIS)

Hora, Stephen

2002-09-01

The role of sensitivity and uncertainty analyses for radioactive waste disposal assessments is reviewed. The report covers a description of the these concepts were applied in the authorities' review of the safety report SR 97. With regard to the use of expert knowledge, the most significant weakness of SR 97 is absence of any standards, procedures, and even definitions for expert judgment. This situation needs to be dealt with by SKB in the near future as it denigrates the portions of the study that are well done. In developing expert judgment processes, SSI should ensure that SKB creates procedures that guarantee traceability and transparency. This will become very important as the repository system matures and receives greater public scrutiny. Both in the area of scenario creation and expert judgement, there are processes that have gained international acceptance. It would be in the best interest of SKB, and the public, to adhere these accepted approaches

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

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)

A study of internal dosimetry of Am-241 and Sr-90 by dismantling of a nuclear installation; Eine Fallstudie zur internen Dosimetrie von Am-241 und Sr-90 bei Rueckbau einer kerntechnischen Anlage

Energy Technology Data Exchange (ETDEWEB)

Froning, M.; Hill, P. [Forschungszentrum Juelich GmbH (Germany). Geschaeftsbereich Sicherheit und Strahlenschutz

2016-07-01

During dismantling operation in former nuclear facility routine incorporation monitoring had been part of the safety measures. For an occupational radiation worker positive measurements results for {sup 241}Am, {sup 90}Sr and {sup 137}Cs were obtained after the end of the working period. Follow up monitoring had been performed assessing urine and faeces samples for {sup 241}Am and {sup 90}Sr as well as in-vivo measurements for {sup 137}Cs. Ingestion could be proven as incorporation path. The internal dose assessment according to GMBl 2007{sup [1]} finally yielded internal dose at 13 μSv.

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)

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...

Development of 87Sr/86Sr maps as targeted strategy to support wine quality.

Science.gov (United States)

Durante, Caterina; Bertacchini, Lucia; Cocchi, Marina; Manzini, Daniela; Marchetti, Andrea; Rossi, Maria Cecilia; Sighinolfi, Simona; Tassi, Lorenzo

2018-07-30

This study summarizes the results obtained from a systematic and long-term project aimed at the development of tools to assess the provenance of food in the oenological sector. In particular, 87 Sr/ 86 Sr isotope ratios were measured on statistically representative set of soils, vine branches and wines sampled in the production district of Modena, worldwide known for the Lambrusco wines production. The obtained data were used to build strontium isotopic maps able to objectively support the Lambrusco PDO wines origin as well as other products of the Modena district. Finally, a strong relationship was found between the 87 Sr/ 86 Sr isotope ratios of soils and vine branches on a large scale, highlighting and confirming once more the idea that plants can also represent an optimal sampling device to support geographical traceability. Copyright © 2018 Elsevier Ltd. All rights reserved.

The adsorption of Cs+, Sr2+ and Ni2+ on bitumen: a mechanistic model

International Nuclear Information System (INIS)

Loon, L.R. Van; Kopajtic, Z.

1991-01-01

The adsorption of radionuclides on the waste matrix is a positive effect and contributes to the retardation of released radionuclides migrating to the geo-and biosphere. For the safety assessment studies, it is important to know whether or not radionuclides do adsorb on the waste matrix. In the present work the adsorption of 134 Cs + , 85 Sr 2+ and 63 Ni 2+ on bitumen was studied as a function of the pH and ionic strength of the equilibrium solution. Bitumen emulsions with well defined surfaces were used. The surface of bitumen is negatively charged due to the deprotonation of weak acid carboxyl groups at the interface. The functional group density amounts to 1.37.10 18 groups/m 2 and their deprotonation behaviour can be well described by the 'Ionizable Surface Group' model. Cs + , Sr 2+ and Ni 2+ adsorb on the surface by three different processes, i.e. ion exchange, outer sphere complexation and inner sphere surface complexation respectively. The adsorption depends on the pH and the ionic strength of the contact solution. Under near field conditions, Cs + and Sr 2+ do not adsorb on the bitumen due to the competition with Na + , K + and Ca 2+ present in the cement pore water in contact with the bitumen. Ni 2+ adsorption can also be neglected because the formation of neutral and anionic hydroxo complexes in solution competes strongly with the adsorption reaction. Other hydrolysable radionuclides of interest are expected to behave similarly to Ni 2+ . The main conclusion of the study is that the adsorption of radionuclides under near field conditions is expected to be very low. Consequently, this process need not to be considered in safety assessment studies. (author) figs., tabs., 30 refs

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 the Mechanisms for Sr-90 and TRU Removal from Complexant-Containing Tank Wastes at Hanford

International Nuclear Information System (INIS)

Hallen, Richard T.; Geeting, John GH; Lilga, Michael A.; Hart, Todd R.; Hoopes, Francis V.

2005-01-01

Small-scale tests (∼20 mL) were conducted with samples from Hanford underground storage tanks AN-102 and AN-107 to assess the mechanisms for removing Sr-90 and transuranics (TRU) from the liquid (supernatant) portion of the waste. The Sr-90 and TRU must be removed (decontaminated), in addition to Cs-137 and the entrained solids, before the supernatant can be disposed of as low-activity waste. Experiments were conducted with various reagents and modified Sr/TRU removal process conditions to more fully understand the reaction mechanisms. The optimized treatment conditions--no added hydroxide, addition of Sr (0.02M target concentration) followed by sodium permanganate (0.02M target concentration) with mixing at ambient temperature--were used as a reference for comparison. The waste was initially two orders of magnitude undersaturated with Sr; the addition of nonradioactive Sr(NO?) ? saturated the supernatant, resulting in isotopic dilution and precipitation of Sr-90 as SrCO?. The reaction chemistry of Mn species relevant to the mechanism of TRU removal by permanganate treatment was evaluated, along with the importance of various mechanisms for decontamination, such as precipitation, absorption, ligand exchange, and oxidation of organic complexants. For TRU removal, permanganate addition generally gave the highest DF. The addition of Mn of lower oxidation states (II, IV, and VI) also resulted in good TRU removal, as did complexant oxidation with periodate and addition of Zr(IV) for ligand exchange. These results suggest that permanganate treatment leads to TRU removal by multiple routes

Can information surety be assessed with high confidence?

International Nuclear Information System (INIS)

Lim, J.J.; Fletcher, S.K.; Halbgewachs, R.D.; Jansma, R.M.; Sands, P.D.; Watterberg, P.A.; Wyss, G.D.

1994-01-01

Several basic reasons are given to support the position that an integrated, systems methodology entailing probabilistic assessment offers the best means for addressing the problems in software safety. The recognized hard problems in software safety, or safety per se, and some of the techniques for hazard identification and analysis are then discussed relative to their specific strengths and limitations. The paper notes that it is the combination of techniques that will lead to safer systems, and that more experience, examples, and applications of techniques are needed to understand the limits to which software safety can be assessed. Lastly, some on-going project work at Sandia National Laboratories on developing a solution methodology is presented

Health and Safety Laboratory environmental quarterly. Final tabulation of monthly /sup 90/Sr fallout data: 1954--1976

Energy Technology Data Exchange (ETDEWEB)

1977-10-01

This report presents the monthly /sup 90/Sr deposition data derived from a global network of stations started in 1954. This program was carried out to assess the distribution patterns and inventory the amount of fallout of radionuclies from atmospheric nuclear tests. In 1976, monthly deposition rates had diminished to the point where measurable levels of /sup 90/Sr were rarely observed.

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)

Comparative analysis of safety related site characteristics

International Nuclear Information System (INIS)

Andersson, Johan

2010-12-01

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Comparative analysis of safety related site characteristics

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan (ed.)

2010-12-15

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

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.

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

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...

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.

SR 97 - Waste, repository design and sites. Background report to SR 97 SKB

International Nuclear Information System (INIS)

1999-10-01

SR 97 is a comprehensive analysis of long-term safety of a deep repository for spent nuclear fuel. The repository is assumed to be designed according to the KBS-3 method. Assessments are performed in SR 97 for three fictitious sites: Aberg, Beberg and Ceberg. One premise is that data used for assessment of the fictitious sites are to be taken from sites that have previously been investigated. The spent nuclear fuel is enclosed in copper canisters with an insert of cast iron. The canisters are emplaced in bored holes in the floor of the deposition tunnels. Around each canister, bentonite blocks are stacked which, after absorbing water and swelling, will isolate the canister from groundwater, hold the canister in place and retard transport of radionuclides from the canister to the surrounding rock. The spent nuclear fuel will emit heat for a long time, due to the decay heat. The maximum permissible temperature on the canister surface has been chosen at 100 deg C. The spacing between the deposition holes and between the deposition tunnels is adjusted site-specifically to meet this requirement. The thermal properties of the rock and the buffer material are of importance for how closely the deposition holes and tunnels can be spaced. After deposition, the deposition tunnels are backfilled with a mixture of bentonite and crushed rock. SR 97 examines above all the consequences of various scenarios and the handling of various types of uncertainties. The different repository sites illustrate normal properties for Swedish bedrock which are of importance for safety. To facilitate the work, the repositories on the three sites are configured as similarly as possible, which means for example that they are located at roughly the same depth and are fitted into the bedrock in a relatively similar fashion. Apart from the siting of a repository for spent nuclear fuel, the site may need to house a separate repository for other long-lived waste. This possibility has been considered in

SR 97 - Waste, repository design and sites. Background report to SR 97 SKB

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-10-01

SR 97 is a comprehensive analysis of long-term safety of a deep repository for spent nuclear fuel. The repository is assumed to be designed according to the KBS-3 method. Assessments are performed in SR 97 for three fictitious sites: Aberg, Beberg and Ceberg. One premise is that data used for assessment of the fictitious sites are to be taken from sites that have previously been investigated. The spent nuclear fuel is enclosed in copper canisters with an insert of cast iron. The canisters are emplaced in bored holes in the floor of the deposition tunnels. Around each canister, bentonite blocks are stacked which, after absorbing water and swelling, will isolate the canister from groundwater, hold the canister in place and retard transport of radionuclides from the canister to the surrounding rock. The spent nuclear fuel will emit heat fora long time, due to the decay heat. The maximum permissible temperature on the canister surface has been chosen at 100 deg C. The spacing between the deposition holes and between the deposition tunnels is adjusted site-specifically to meet this requirement. The thermal properties of the rock and the buffer material are of importance for how closely the deposition holes and tunnels can be spaced. After deposition, the deposition tunnels are backfilled with a mixture of bentonite and crushed rock. SR 97 examines above all the consequences of various scenarios and the handling of various types of uncertainties. The different repository sites illustrate normal properties for Swedish bedrock which are of importance for safety. To facilitate the work, the repositories on the three sites are configured as similarly as possible, which means for example that they are located at roughly the same depth and are fitted into the bedrock in a relatively similar fashion. Apart from the siting of a repository for spent nuclear fuel, the site may need to house a separate repository for other long-lived waste. This possibility has been considered in

Risk perception, risk management and safety assessment: what can governments do to increase public confidence in their vaccine system?

Science.gov (United States)

MacDonald, Noni E; Smith, Jennifer; Appleton, Mary

2012-09-01

For decades vaccine program managers and governments have devoted many resources to addressing public vaccine concerns, vaccine risk perception, risk management and safety assessment. Despite ever growing evidence that vaccines are safe and effective, public concerns continue. Education and evidence based scientific messages have not ended concerns. How can governments and programs more effectively address the public's vaccine concerns and increase confidence in the vaccine safety system? Vaccination hesitation has been attributed to concerns about vaccine safety, perceptions of high vaccine risks and low disease risk and consequences. Even when the public believes vaccines are important for protection many still have concerns about vaccine safety. This overview explores how heuristics affect public perception of vaccines and vaccine safety, how the public finds and uses vaccine information, and then proposes strategies for changes in the approach to vaccine safety communications. Facts and evidence confirming the safety of vaccines are not enough. Vaccine beliefs and behaviours must be shaped. This will require a shift in the what, when, how and why of vaccine risk and benefit communication content and practice. A change to a behavioural change strategy such as the WHO COMBI program that has been applied to disease eradication efforts is suggested. Copyright © 2011. Published by Elsevier Ltd.. All rights reserved.

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

SR 97. Processes in the repository evolution. Background report to SR 97

Energy Technology Data Exchange (ETDEWEB)

Hedin, A. [ed.

1999-11-01

This report describes, in a comprehensive and coherent fashion, all identified internal processes of importance for the post-closure evolution and safety of a KBS-3 repository for spent nuclear fuel. The report has been written to be used in the SR 97 project, which has limited the time available for its preparation. Differences in the level of detail in descriptions of different processes do not always reflect differences in the significance of the processes. Discussions of different types of uncertainties could in many cases be broadened and deepened, and the stylistic quality could sometimes be improved. Like other background material for the safety assessments, the process report is also expected to require revision as site-specific conditions are progressively clarified. Today's version of the process report is therefore the firstversion of a report that will be revised prior to every safety report. The intention is to perform the first revision of the report after scrutiny of SR 97. The report describes the internal processes which over time lead to changes in a KBS-3 repository for spent nuclear fuel. The context of the material in the report is described in SR 97 Main Report and briefly entails the following: The repository has been divided into four subsystems: fuel/cavity, cast iron insert/copper canister, buffer/backfill and geosphere. A number of processes of importance for the post-closure evolution of the repository have been identified within each subsystem. This has been done with the aid of material in the so-called interaction matrices previously developed by SKB. The processes have been divided into the categories thermal, hydraulic, mechanical and chemical. Furthermore, there are processes related to radiation and radionuclide transport. The identified processes are documented in this report. Each subsystem has its own chapter, and each chapter is divided into radiation related, thermal, hydraulic, mechanical and chemical processes as well

SR 97. Processes in the repository evolution. Background report to SR 97

International Nuclear Information System (INIS)

Hedin, A.

1999-11-01

This report describes, in a comprehensive and coherent fashion, all identified internal processes of importance for the post-closure evolution and safety of a KBS-3 repository for spent nuclear fuel. The report has been written to be used in the SR 97 project, which has limited the time available for its preparation. Differences in the level of detail in descriptions of different processes do not always reflect differences in the significance of the processes. Discussions of different types of uncertainties could in many cases be broadened and deepened, and the stylistic quality could sometimes be improved. Like other background material for the safety assessments, the process report is also expected to require revision as site-specific conditions are progressively clarified. Today's version of the process report is therefore the first version of a report that will be revised prior to every safety report. The intention is to perform the first revision of the report after scrutiny of SR 97. The report describes the internal processes which over time lead to changes in a KBS-3 repository for spent nuclear fuel. The context of the material in the report is described in SR 97 Main Report and briefly entails the following: The repository has been divided into four subsystems: fuel/cavity, cast iron insert/copper canister, buffer/backfill and geosphere. A number of processes of importance for the post-closure evolution of the repository have been identified within each subsystem. This has been done with the aid of material in the so-called interaction matrices previously developed by SKB. The processes have been divided into the categories thermal, hydraulic, mechanical and chemical. Furthermore, there are processes related to radiation and radionuclide transport. The identified processes are documented in this report. Each subsystem has its own chapter, and each chapter is divided into radiation related, thermal, hydraulic, mechanical and chemical processes as well as

SR 97. Processes in the repository evolution. Background report to SR 97

International Nuclear Information System (INIS)

Hedin, A.

1999-11-01

This report describes, in a comprehensive and coherent fashion, all identified internal processes of importance for the post-closure evolution and safety of a KBS-3 repository for spent nuclear fuel. The report has been written to be used in the SR 97 project, which has limited the time available for its preparation. Differences in the level of detail in descriptions of different processes do not always reflect differences in the significance of the processes. Discussions of different types of uncertainties could in many cases be broadened and deepened, and the stylistic quality could sometimes be improved. Like other background material for the safety assessments, the process report is also expected to require revision as site-specific conditions are progressively clarified. Today's version of the process report is therefore the first version of a report that will be revised prior to every safety report. The intention is to perform the first revision of the report after scrutiny of SR 97. The report describes the internal processes which over time lead to changes in a KBS-3 repository for spent nuclear fuel. The context of the material in the report is described in SR 97 Main Report and briefly entails the following: The repository has been divided into four subsystems: fuel/cavity, cast iron insert/copper canister, buffer/backfill and geosphere. A number of processes of importance for the post-closure evolution of the repository have been identified within each subsystem. This has been done with the aid of material in the so-called interaction matrices previously developed by SKB. The processes have been divided into the categories thermal, hydraulic, mechanical and chemical. Furthermore, there are processes related to radiation and radionuclide transport. The identified processes are documented in this report. Each subsystem has its own chapter, and each chapter is divided into radiation related, thermal, hydraulic, mechanical and chemical processes as well as

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)

30 CFR 56.4402 - Safety can use.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Safety can use. 56.4402 Section 56.4402 Mineral... and Combustible Liquids and Gases § 56.4402 Safety can use. Small quantities of flammable liquids drawn from storage shall be kept in safety cans labeled to indicate the contents. ...

30 CFR 57.4402 - Safety can use.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Safety can use. 57.4402 Section 57.4402 Mineral... Flammable and Combustible Liquids and Gases § 57.4402 Safety can use. Small quantities of flammable liquids drawn from storage shall be kept in safety cans labeled to indicate the contents. ...

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

Is the Modern Marine 87Sr/86Sr Cycle Balanced?

Science.gov (United States)

Peucker-Ehrenbrink, B.

2017-12-01

The marine 87Sr/86Sr record is one of the best-reconstructed isotope records with thousands of high quality measurements spanning the past 800 million years. It records a global signal of tectonic, biotic and climatic processes on Earth. Yet despite decades of research we still do not know whether the current marine Sr budget is in steady state. Studies of the marine 88Sr/86Sr record indicate that sources and sinks do not balance. The magnitude and isotope composition of the terrestrial inputs are being debated, and the magnitude and temporal variability of unradiogenic contributions are not well constrained. Here I provide a revised assessment of all continental sources of Sr to the ocean, including river runoff, submarine groundwater discharge (Beck et al., 2013), dissolution of riverine suspended matter in seawater and dissolution of volcanic ash deposited on the ocean (Jones et al., 2012). I contrast continental sources of Sr with estimates of marine sources of Sr to seawater, specifically high- and low-temperature submarine hydrothermal fluids, as well as diffusive diagenetic fluxes. Best current data imply that unradiogenic submarine hydrothermal inputs to seawater are insufficient to balance the flux of radiogenic continental Sr. The revised assessment of riverine contributions is based on Sr data for almost 230 rivers, an increasing amount of time-series data for such rivers, as well as river discharge and sediment flux data for more than 2000 rivers. Regional sampling biases have been corrected with the aid of digital bedrock maps, specifically along the western margin of North America, East Africa and the large drainage region of Arabia, India and SE Asia. Significant uncertainty in the chemical and isotopic compositions of runoff from Greenland and East Africa remains. The main uncertainty in the budget, however, is related to the possibility that modern rivers do not represent the pre-anthropogenic (natural) state of continental runoff (e.g. Ganges

Environmental levels of 239+240Pu and 90Sr for internal radiation exposure assessment

International Nuclear Information System (INIS)

Anand, S.J.S.; Khandekar, R.N.; Krishnamoorthy, T.M.

1995-01-01

Measurements have been carried out on the concentration of low levels of long-lived isotopes of 239+240 Pu and 90 Sr in the environmental materials such as atmospheric particulates, drinking water and food. The estimation of daily intake of these isotopes through inhalation and ingestion is a pre-requisite for the assessment of internal exposure. This paper presents temporal distribution of 239+240 Pu and 90 Sr in rain water, drinking water and total diet samples collected at Trombay site. The annual committed effective dose due to 90 Sr through inhalation and diet to the population of Bombay has been estimated to be 0.06 nSv/y and 0.48 μSv/y, respectively, and the same for 239+240 Pu is 1.3 nSv/y and 0.9 nSv/y, respectively. The data is discussed in relation to previous years' values to assess for any significant increase. (author). 9 refs., 3 figs., 2 tabs

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

International Nuclear Information System (INIS)

Jones, G.

2013-01-01

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

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

Evaluation of SKB's report 'Deep repository for spent nuclear fuel: SR 97 - Post-closure safety', Focusing on the assessment of transport processes in the geosphere

Energy Technology Data Exchange (ETDEWEB)

Woerman, A.; Shulan Xu [Uppsala Univ. (Sweden). Dept. of Geoscience

2000-12-01

This report describes a critical review of the safety assessment performed on the final repository for nuclear waste in Sweden that is proposed by SKB in 'Deep Repository for Spent Nuclear Fuel: SR 97 - Post-closure Safety'. The review was requested by the Swedish Nuclear Power Inspectorate (SKI). The waste repository consists of several barriers that work together with the purpose of delaying radionuclide migration and reducing the activity that eventually affects the biosphere. A main criticism is the lack of a formal risk analysis and uncertainties in several analyses that make it difficult to comprehend the overall risk of the repository. A formal risk analysis should comprise a probabilistic treatment of all components included in the system. This is not the case in the SKB's report since the probabilistic analyses are limited only to certain aspects. The use of conservative model parameters are not a substitute for risk analysis nor can they compensate for possible model biases. Bias can be expected in most of the existing models of radionuclide migration in fractured bedrock. SKB should present a clear comparison on the importance of the different barrier components (uranium-dioxide matrix, copper canister, buffer and bedrock) on the retardation of radionuclides. It is unclear as to what extent the capacity of the bedrock to retain migrating radionuclides is critical to the capacity of the repository. A large part of the SR 97 report is focused on retardation processes in bedrock and a reader can interpret this as the technical weight given on retardation in the bedrock. However, with the present state of knowledge, it is our opinion that we cannot with an acceptable degree of accuracy predict the radionuclide transport in bedrock or quantify risk levels associated with radioactivity in the biosphere. There are large uncertainties concerning the way by which sorption processes should be formulated and the impact of colloids on the transport

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)

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

Like other nuclear utilities, Ontario Hydro, as part of its risk management activities, continually assesses the safety of its nuclear operations. In addition, new regulatory requirements are being applied to the older nuclear power plants. Both of these result in proposed plant modifications designed to reduce the risk to the public. However, modifications to an operating plant can have serious economic effects, and the resources, both financial and personnel, required for the implementation of these modifications are limited. Thus, all potential benefits and effects of a proposed modification must be thoroughly investigated to judge whether the modification is beneficial. Ontario Hydro has begun to use comprehensive value-impact assessments, utilizing plant-specific probabilistic risk assessments (PRAs), as tools to provide an informed basis for judgments on the benefit of safety-related modifications. The results from value-impact assessments can also be used to prioritize the implementation of these modifications

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!

Assessing electron beam sensitivity for SrTiO{sub 3} and La{sub 0.7}Sr{sub 0.3}MnO{sub 3} using electron energy loss spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Nord, Magnus, E-mail: magnunor@gmail.com [Department of Physics, NTNU, Trondheim (Norway); Vullum, Per Erik [Department of Physics, NTNU, Trondheim (Norway); Materials and Chemistry, SINTEF, Trondheim (Norway); Hallsteinsen, Ingrid; Tybell, Thomas [Department of Electronics and Telecommunications, NTNU, Trondheim (Norway); Holmestad, Randi [Department of Physics, NTNU, Trondheim (Norway)

2016-10-15

Thresholds for beam damage have been assessed for La{sub 0.7}Sr{sub 0.3}MnO{sub 3} and SrTiO{sub 3} as a function of electron probe current and exposure time at 80 and 200 kV acceleration voltage. The materials were exposed to an intense electron probe by aberration corrected scanning transmission electron microscopy (STEM) with simultaneous acquisition of electron energy loss spectroscopy (EELS) data. Electron beam damage was identified by changes of the core loss fine structure after quantification by a refined and improved model based approach. At 200 kV acceleration voltage, damage in SrTiO{sub 3} was identified by changes both in the EEL fine structure and by contrast changes in the STEM images. However, the changes in the STEM image contrast as introduced by minor damage can be difficult to detect under several common experimental conditions. No damage was observed in SrTiO{sub 3} at 80 kV acceleration voltage, independent of probe current and exposure time. In La{sub 0.7}Sr{sub 0.3}MnO{sub 3}, beam damage was observed at both 80 and 200 kV acceleration voltages. This damage was observed by large changes in the EEL fine structure, but not by any detectable changes in the STEM images. The typical method to validate if damage has been introduced during acquisitions is to compare STEM images prior to and after spectroscopy. Quantifications in this work show that this method possibly can result in misinterpretation of beam damage as changes of material properties. - Highlights: • We studied the effects of a TEM electron beam on a perovskite heterostructure. • Using an improved ELNES quantification method, subtle changes could be observed. • On LSMO changes were observed in the ELNES, but none in the STEM-HAADF. • For STO changes were observed in both ELNES and STEM-HAADF. • This shows beam damage can be misinterpreted as material properties.

Can one observe by μ SR the transition from uncorrelated to correlated spin fluctuations? Example: Nd1.4Ce0.2Sr0.4CuO4-δ

International Nuclear Information System (INIS)

Pinkpank, M.; Amato, A.; Gygax, F.N.; Schenck, A.; Henggeler, W.; Fischer, P.

1997-01-01

μSR-measurements in ZF and LF on Nd 1.4 Ce 0.2 Sr 0.4 CuO 4-δ show a sharp increase of the depolarisation rate (λ) below ∼ 2K. This increase can be explained by the transition from uncorrelated to correlated spin fluctuations, which is in agreement with results obtained by neutron scattering

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.

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

NARCIS (Netherlands)

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

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.

Assessing propensity to learn from safety-related events

NARCIS (Netherlands)

Drupsteen, L.; Wybo, J.L.

2015-01-01

Most organisations aim to use experience from the past to improve safety, for instance through learning from safety-related incidents and accidents. Whether an organisation is able to learn successfully can however only be determined afterwards. So far, there are no proactive measures to assess

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...

Experiences with the determination of Sr-89 and Sr-90 using fast methods; Erfahrungen bei der Bestimmung von {sup 89}Sr und {sup 90}Sr mittels Schnellmethoden

Energy Technology Data Exchange (ETDEWEB)

Kowalik, C.; Fueger, J. [Thueringer Landesanstalt fuer Umwelt und Geologie, Jena (Germany). Landesmessstelle fuer Umweltradioaktivaet

2014-01-20

Quick methods of the measurement of {sup 89}Sr and {sup 90}Sr have a great importance in the supervision of the environmental radioactivity. It is necessary to receive in short time dependable analytical data to be able to carry out suitable assessments or to give recommendations. The aim of the investigations was to be guaranteed the demands for these methods (test preparation, measurement and evaluation). The use of the solid phase extraction by means of commercial Sr Resin trademark columns (4.4' (5')-Di-tert-butylcyclohexanol-18-kronen-6-aether) (Triskem) to the radiochemical separation of the Sr isotopes was suitable. The measurements occurred to the FHT 770 T12 - Multi Low Level Alpha/Beta Sample Counter (Thermo Scientific). The results contain the summary activities of all available Sr isotopes, as for example {sup 89}Sr and {sup 90}Sr. The calculations of the single activities occur about the mathematical algorithm of the linear development on the basis of the works of G. Kanisch. The first results show, this method is suitable for the analysis of {sup 89}Sr and {sup 90}Sr and is used therefore in future in Thuringia.

Official News relating to CERN Safety Rules

CERN Multimedia

HSE Unit

2015-01-01

The CERN Safety Rules listed below have been published on the HSE website (see here) and entered into force on the 9 June 2015:   Safety Regulation SR-M “Mechanical equipment”: http://cern.ch/safety-rules/SR-M_ENv2.htm; this SR-M (version 2) cancels and replaces SR-M (version 1) and the corresponding provisions of General Safety Instruction GSI-M3 “Special Equipment” (version 1).   General Safety Instruction GSI-M-1 “Lifting equipment and accessories”: http://cern.ch/safety-rules/GSI-M-1_ENv2.htm; this GSI-M-1 (version 2) cancels and replaces GSI-M1 (version 1). Specific Safety Instruction SSI-M-1-1 “Slings and lifting chains”: http://cern.ch/safety-rules/SSI-M-1-1_EN.htm; Specific Safety Instruction SSI-M-1-2 “Cranes, bridge cranes, gantry cranes and power-driven hoists”: http://cern.ch/safety-rules/SSI-M-1-2_EN.htm; Specific Safety Instruction SSI-M-1-3 “Non-f...

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

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

Electronic properties and surface reactivity of SrO-terminated SrTiO3 and SrO-terminated iron-doped SrTiO3.

Science.gov (United States)

Staykov, Aleksandar; Tellez, Helena; Druce, John; Wu, Ji; Ishihara, Tatsumi; Kilner, John

2018-01-01

Surface reactivity and near-surface electronic properties of SrO-terminated SrTiO 3 and iron doped SrTiO 3 were studied with first principle methods. We have investigated the density of states (DOS) of bulk SrTiO 3 and compared it to DOS of iron-doped SrTiO 3 with different oxidation states of iron corresponding to varying oxygen vacancy content within the bulk material. The obtained bulk DOS was compared to near-surface DOS, i.e. surface states, for both SrO-terminated surface of SrTiO 3 and iron-doped SrTiO 3 . Electron density plots and electron density distribution through the entire slab models were investigated in order to understand the origin of surface electrons that can participate in oxygen reduction reaction. Furthermore, we have compared oxygen reduction reactions at elevated temperatures for SrO surfaces with and without oxygen vacancies. Our calculations demonstrate that the conduction band, which is formed mainly by the d-states of Ti, and Fe-induced states within the band gap of SrTiO 3 , are accessible only on TiO 2 terminated SrTiO 3 surface while the SrO-terminated surface introduces a tunneling barrier for the electrons populating the conductance band. First principle molecular dynamics demonstrated that at elevated temperatures the surface oxygen vacancies are essential for the oxygen reduction reaction.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Correlation between safety climate and contractor safety assessment programs in construction.

Science.gov (United States)

Sparer, Emily H; Murphy, Lauren A; Taylor, Kathryn M; Dennerlein, Jack T

2013-12-01

Contractor safety assessment programs (CSAPs) measure safety performance by integrating multiple data sources together; however, the relationship between these measures of safety performance and safety climate within the construction industry is unknown. Four hundred and one construction workers employed by 68 companies on 26 sites and 11 safety managers employed by 11 companies completed brief surveys containing a nine-item safety climate scale developed for the construction industry. CSAP scores from ConstructSecure, Inc., an online CSAP database, classified these 68 companies as high or low scorers, with the median score of the sample population as the threshold. Spearman rank correlations evaluated the association between the CSAP score and the safety climate score at the individual level, as well as with various grouping methodologies. In addition, Spearman correlations evaluated the comparison between manager-assessed safety climate and worker-assessed safety climate. There were no statistically significant differences between safety climate scores reported by workers in the high and low CSAP groups. There were, at best, weak correlations between workers' safety climate scores and the company CSAP scores, with marginal statistical significance with two groupings of the data. There were also no significant differences between the manager-assessed safety climate and the worker-assessed safety climate scores. A CSAP safety performance score does not appear to capture safety climate, as measured in this study. The nature of safety climate in construction is complex, which may be reflective of the challenges in measuring safety climate within this industry. Am. J. Ind. Med. 56:1463-1472, 2013. © 2013 Wiley Periodicals, Inc. © 2013 Wiley Periodicals, Inc.

Buffer erosion: An overview of concepts and potential safety consequences

International Nuclear Information System (INIS)

Apted, Michael J.; Arthur, Randy; Bennett, David; Savage, David; Saellfors, Goeran; Wennerstroem, Haakan

2010-11-01

In its safety analysis SR-Can, SKB reported preliminary results and conclusions on the mechanisms of bentonite colloid formation and stability, with a rough estimate of the consequences of loss of bentonite buffer by erosion. With the review of SR-Can the authorities (SKI and SSI) commented that erosion of the buffer had the greatest safety significance, that the understanding of the mechanisms of buffer erosion was inadequate, and that more work would be required to arrive at robust estimates of the extent and impacts of buffer erosion. After the SR-Can report, SKB started a two-year research project on buffer erosion. The results from this two-year project have been reported in several SKB technical reports. SSM started this project to build up its own competence in the related scientific areas by a preliminary evaluation of SKB's research results

Buffer erosion: An overview of concepts and potential safety consequences

Energy Technology Data Exchange (ETDEWEB)

Apted, Michael J.; Arthur, Randy (INTERA Incorporated, Denver, CO (United States)); Bennett, David (TerraSalus Limited, Rutland (United Kingdom)); Savage, David (Savage Earth Associates Limited, Bournemouth (United Kingdom)); Saellfors, Goeran (GeoForce AB, Billdal (Sweden)); Wennerstroem, Haakan (Dept. of Chemistry, Lund Univ., Lund (Sweden))

2010-11-15

In its safety analysis SR-Can, SKB reported preliminary results and conclusions on the mechanisms of bentonite colloid formation and stability, with a rough estimate of the consequences of loss of bentonite buffer by erosion. With the review of SR-Can the authorities (SKI and SSI) commented that erosion of the buffer had the greatest safety significance, that the understanding of the mechanisms of buffer erosion was inadequate, and that more work would be required to arrive at robust estimates of the extent and impacts of buffer erosion. After the SR-Can report, SKB started a two-year research project on buffer erosion. The results from this two-year project have been reported in several SKB technical reports. SSM started this project to build up its own competence in the related scientific areas by a preliminary evaluation of SKB's research results

The development of safety assessment technology for the radwaste disposal

International Nuclear Information System (INIS)

Han, Kyong Won; Cho, Won Jin; Lee, Han Soo; Lee, Jai Wan; Park, Chung Kyun; Lee, Myun Joo; Cho, Young Hwan; Choi, Heui Joo; Lee, Youn Myoung; Park, Hee Sung

1989-02-01

This report is composed of three parts. In part I, an improved radwaste disposal safety assessment code named SADROCM, was developed by upgrading the existing SADROC. A numerical approach was adopted for the simulation of diffusion into rock pore and advection in the fracture. Also quantification of resaturation time in repository was obtained by introducing theoretical resaturation model and groundwater flow model. In radiation dose model the calculated dose by existing model was compared with that by LIMCAL code. Part II resulted in several findings. Regarding the leaching of radionuclide, two steps were observed. It was identified that in the initial step, radionuclide leaching occurred at the surface of solidified waste, and then leaching proceeded by dissolution and diffusion of radionuclide from inside of the solidified waste.The results of the sorption experiments on Cs, Co, Sr for the rock samples from four regions, showed that Kd value of F-site quartz-porphery was the lowest, and that of B-site Rhyolitic tuff the highest. E-site granite, M-site granite gneiss showed medium sorption capacities. The Kd values of each radionuclide were in the following order: Cs > Co > Sr . For the radionuclide migration through fractured rock, examination of hydraulic dispersion model and channeling model was carried out to simulate the dispersion phenomena through the fractured rock. In part III, optimum model was established which allows mathematical prediction of the container corrosion rate and leach rate.Based on this model, the equations were obtained to predict corrosion rate and radionuclide leach rate. A good agreement of these equations and existing experimental data was considered to be an indication that it can be used effectively for the estimation of long-term leaching. (Author)

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.

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)

Review of SR-Can: Evaluation of SKB's handling of spent fuel performance, radionuclide chemistry and geosphere transport parameters. External review contribution in support of SKI's and SSI's review of SR-Can

Energy Technology Data Exchange (ETDEWEB)

Stenhouse, Mike (Monitor Scientific LLC, Denver, CO (US)); Jegou, Christophe (Commissariat a l' Energie Atomique (CEA) (FR)); Brown, Paul (Geochem Australia (AU)); Meinrath, Guenther (RER Consultants, Passau (DE)); Nitsche, Heino (Univ. of California, Berkeley (US)); Ekberg, Christian (Chalmers University of Technology (SE))

2008-03-15

SR-Can covers the containment phase of the KBS-3 barriers as well as the consequences of releases of radionuclides to the rock and eventually the biosphere (after complete containment within fuel canisters has partially failed). The aim of this report is to provide a range of review comments with respect to those parameters related to spent fuel performance as well as radionuclide chemistry and transport. These parameter values are used in the quantification of consequences due to release of radionuclides from potentially leaking canisters. The report does not cover modelling approaches used for quantification of consequences. However, modelling used to derive parameter values is to some extent addressed (such as calculation of maximum radionuclide concentration due to formation of solubility limiting phases). The following are the key highlights and comments generated in the course of the review: Inconsistencies exist between recommendations provided in technical reports and those quoted in the Data Report. One of the reasons for such inconsistencies has been the timing of different pieces of research. It is hoped that the timing of contributions to SR-Site will be such that these inconsistencies can be avoided. Sensitivity analyses need to be carried out and reported in a number of areas to support some of the assumptions or decisions made in the assessment calculations. The likelihood is that SKB has performed many of the sensitivity analyses identified in different parts of this report, but these need to be reported, preferably to complement the recommendations provided

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)

Fewer can be More: Nuclear Safety and Security Culture Self-Assessment in the Hungarian Public Ltd. for Radioactive Waste Management

International Nuclear Information System (INIS)

Horváth, K.; Solymosi, M.; Vass, G.

2016-01-01

The Hungarian regulator and operators show strong commitment towards robust nuclear safety and security culture. The paper discusses the evolution and the basis of the regulation of Hungarian safety and security culture. Because of security considerations nuclear safety incidents have always received and for sure will receive more publicity than malicious acts. That is probably the main reason behind that mostly nuclear safety incidents influence the common beliefs. This kind of primacy is noticeable as well in regulations and also in practice. Although there is a strong connection nuclear safety and security culture, their relationship has not been researched for a long time. The paper also presents an already achieved, combined nuclear safety and security culture survey type assessment. Survey is a well known type of organizational culture self assessment. The applied methods, relationship between these two cultures and of course some difficulties of the process are summarized. The presented method is appropriate to combine different guidance and characteristics to measure different attitude in a single survey. The method in practice is shown through the nuclear safety and security culture assessment conducted at Hungarian Public Ltd. Of Radioactive Waste Management. (author)

Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

International Nuclear Information System (INIS)

1986-11-01

Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

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

Groundwater chemistry around a repository for spent nuclear fuel over a glacial cycle. Evaluation for SR-Can

Energy Technology Data Exchange (ETDEWEB)

Auque, L.F.; Gimeno, M.J.; Gomez, J.B. [University of Zaragoza (Spain); Puigdomenech, I. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Smellie, J. [Conterra AB, Uppsala (Sweden); Tullborg, E.L. [Terralogica AB, Graabo (Sweden)

2007-12-15

, and indicate that far-field groundwaters during the temperate period will not affect negatively the performance of the safety functions of the repository. For the permafrost and glacial periods, groundwater compositions are proposed based on the results from hydrological evaluations performed within the SR-Can project. It is concluded that for permafrost conditions groundwaters in the rock volume surrounding a repository will not affect negatively its performance. However, under glacial conditions meltwaters are expected to penetrate deep into the bedrock and, on the whole, groundwaters would then have such low salinities that it would affect negatively the stability of the bentonite buffer surrounding the canisters in the repository. An analysis of the possibility of penetration of O{sub 2}-rich meltwaters down to repository depths during glacial periods is made based on studies performed within the SR-Can project and elsewhere. It is concluded the inflow of oxygen in single fractures is neutralised by the process of matrix diffusion and dissolution of Fe(II) minerals in the rock matrix. For fracture zones, with water advective times down to repository depths of only a few years, advancement of O{sub 2}-rich waters to repository depth does not occur if variables are cautiously selected. For extreme situations, and given our present understanding of fracture zones, the occurrence of oxidizing conditions at repository depths within large fracture zones can not at present be completely ruled out, but at least they can be avoided in the repository design.

Groundwater chemistry around a repository for spent nuclear fuel over a glacial cycle. Evaluation for SR-Can

International Nuclear Information System (INIS)

Auque, L.F.; Gimeno, M.J.; Gomez, J.B.; Puigdomenech, I.; Smellie, J.; Tullborg, E.L.

2007-12-01

, and indicate that far-field groundwaters during the temperate period will not affect negatively the performance of the safety functions of the repository. For the permafrost and glacial periods, groundwater compositions are proposed based on the results from hydrological evaluations performed within the SR-Can project. It is concluded that for permafrost conditions groundwaters in the rock volume surrounding a repository will not affect negatively its performance. However, under glacial conditions meltwaters are expected to penetrate deep into the bedrock and, on the whole, groundwaters would then have such low salinities that it would affect negatively the stability of the bentonite buffer surrounding the canisters in the repository. An analysis of the possibility of penetration of O 2 -rich meltwaters down to repository depths during glacial periods is made based on studies performed within the SR-Can project and elsewhere. It is concluded the inflow of oxygen in single fractures is neutralised by the process of matrix diffusion and dissolution of Fe(II) minerals in the rock matrix. For fracture zones, with water advective times down to repository depths of only a few years, advancement of O 2 -rich waters to repository depth does not occur if variables are cautiously selected. For extreme situations, and given our present understanding of fracture zones, the occurrence of oxidizing conditions at repository depths within large fracture zones can not at present be completely ruled out, but at least they can be avoided in the repository design

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)

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

International Nuclear Information System (INIS)

Hedin, Allan

2002-12-01

feedback given as to whether further characterisation in this respect is warranted. The total time required for a preliminary safety assessment project is estimated at five months, during which time many of the subtasks would be carried out simultaneously. In parallel to the preliminary safety evaluation, a renewed safety assessment of the KBS-3 method, called SR-MET, will be reported, where new developments regarding analysis methodology and barrier performance will be accounted for. Much of the methodology presented in that report will require more detailed site data than will be available from the candidate sites at the required point in time and such data will be taken from other, previously investigated, sites. SR-MET can be regarded as a template for the safety report SR-SITE, which will be based on data from the complete site investigation. The results of the analyses presented in SR-MET will also provide feedback to continued site investigations and to repository engineering. An example of this would be the SR-MET analyses of the long-term evolution of different backfill materials for various external conditions, the results of which will provide feedback to repository engineering for the site specific choice of a suitable material

Probabilistic assessment methods as a tool for developing nations to make safety decisions

International Nuclear Information System (INIS)

Gumley, P.; Inamdar, S.V.

1985-01-01

This paper advocates the use of probabilistic safety assessment methods in making safety decisions. It discusses the question of adequate safety - what it means to a country buying a nuclear power plant, and how probabilistic safety assessment studies of the reference plant can be used for ensuring this adequate safety. It is proposed that adequate safety means ensuring that the plant would behave, in accident conditions, in a manner similar to the way it is expected to behave were it in the country of origin. For this one needs to know how the plant responds under somewhat altered conditions. These altered conditions can arise from such factors as varying reliability of electrical grids, different manufacturing technology, local systems design and operator capability. In the design of nuclear power plants, the traditional approach to safety has led to the belief that availability and effectiveness of safety systems alone are all that is required to ensure plant safety. This belief can result in design oversights leading to potential problems arising from the power production systems and the service systems. Participation by the buying country in the design of such systems, and understanding the safety implications thereof, can be facilitated by probabilistic safety assessment methods. This philosophy is illustrated in this paper by examples. (author)

78 FR 14912 - International Aviation Safety Assessment (IASA) Program Change

Science.gov (United States)

2013-03-08

... Aviation Safety Assessment (IASA) Program Change AGENCY: Federal Aviation Administration (FAA), DOT. ACTION..., into the U.S., or codeshare with a U.S. air carrier, complies with international aviation safety... subject to that country's aviation safety oversight can serve the United States using its own aircraft or...

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.

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.

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)

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 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

Safety assessment of genetically modified foods

NARCIS (Netherlands)

Kleter, G.A.; Noordam, M.Y.

2016-01-01

The cultivation of genetically modified (GM) crops has steadily increased since their introduction to the market in the mid-1990s. Before these crops can be grown and sold they have to obtain regulatory approval in many countries, the process of which includes a pre-market safety assessment. The

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)

Review. Deep repository for spent nuclear fuel SR 97 - Post-closure safety

International Nuclear Information System (INIS)

Stephansson, Ove

2000-01-01

SKB states that the chosen scenarios provide good coverage of future evolutionary pathways for the deep repository. This is not the case. SKB has not made full use of the established interaction matrices and the new method of THMC diagrams to generate the relevant and important scenarios and to construct the important pathways of variables and processes, either in the established interaction matrices and the presented THMC diagrams. Hence, SKB is demonstrating in SR 97 that they lack a well thought through, sound and solid method to select and evaluate scenarios for the purpose of demonstrating the safety of a deep repository for spent nuclear fuel. The evolution of the system is presented for the components of the repository system (fuel, canister, buffer/backfill, geosphere) and the effects of four different scenarios, but time only enters into the system for discrete events or processes, e.g. description of the relative radiotoxicity and heat decay of the fuel, temperature distribution, iron exchange process, pH in buffer, redox capacity and radionuclear release at the three sites. There is a lack of method and way of describing the evolution of the complete repository system, including the major scenarios, as a function of time. It is essential that SKB is able to: - consider the full range of potential scenarios, - grade the scenarios according to their significance for repository design and performance and safety assessment, - consider whether simple engineering actions could be taken to inhibit the development of adverse scenarios. This cannot be done with the system presented in SR 97, and so SKB do not have a full predictive capability - which is required for the engineering design of such an important and costly structure as a repository. Geoscientific investigation material for three selected sites are presented by SKB in the technical report dealing with waste, repository design and sites. Here a general overview is missing of the geological and rock

Review. Deep repository for spent nuclear fuel SR 97 - Post-closure safety

Energy Technology Data Exchange (ETDEWEB)

Stephansson, Ove [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Civil and Environmental Engineering

2000-12-01

SKB states that the chosen scenarios provide good coverage of future evolutionary pathways for the deep repository. This is not the case. SKB has not made full use of the established interaction matrices and the new method of THMC diagrams to generate the relevant and important scenarios and to construct the important pathways of variables and processes, either in the established interaction matrices and the presented THMC diagrams. Hence, SKB is demonstrating in SR 97 that they lack a well thought through, sound and solid method to select and evaluate scenarios for the purpose of demonstrating the safety of a deep repository for spent nuclear fuel. The evolution of the system is presented for the components of the repository system (fuel, canister, buffer/backfill, geosphere) and the effects of four different scenarios, but time only enters into the system for discrete events or processes, e.g. description of the relative radiotoxicity and heat decay of the fuel, temperature distribution, iron exchange process, pH in buffer, redox capacity and radionuclear release at the three sites. There is a lack of method and way of describing the evolution of the complete repository system, including the major scenarios, as a function of time. It is essential that SKB is able to: - consider the full range of potential scenarios, - grade the scenarios according to their significance for repository design and performance and safety assessment, - consider whether simple engineering actions could be taken to inhibit the development of adverse scenarios. This cannot be done with the system presented in SR 97, and so SKB do not have a full predictive capability - which is required for the engineering design of such an important and costly structure as a repository. Geoscientific investigation material for three selected sites are presented by SKB in the technical report dealing with waste, repository design and sites. Here a general overview is missing of the geological and rock

Economic aspects of risk assessment in chemical safety

Energy Technology Data Exchange (ETDEWEB)

Drummond, M F; Shannon, H S

1986-05-01

This paper considers how the economic aspects of risk assessment in chemical safety can be strengthened. Its main focus is on how economic appraisal techniques, such as cost-benefit and cost-effectiveness analysis, can be adapted to the requirements of the risk-assessment process. Following a discussion of the main methodological issues raised by the use of economic appraisal, illustrated by examples from the health and safety field, a number of practical issues are discussed. These include the consideration of the distribution of costs, effects and benefits, taking account of uncertainty, risk probabilities and public perception, making the appraisal techniques useful to the early stages of the risk-assessment process and structuring the appraisal to permit continuous feedback to the participants in the risk-assessment process. It is concluded that while the way of thinking embodied in economic appraisal is highly relevant to the consideration of choices in chemical safety, the application of these principles in formal analysis of risk reduction procedures presents a more mixed picture. The main suggestions for improvement in the analyses performed are the undertaking of sensitivity analyses of study results to changes in the key assumptions, the presentation of the distribution of costs and benefits by viewpoint, the comparison of health and safety measures in terms of their incremental cost per life-year (or quality-adjusted life-year) gained and the more frequent retrospective review and revision of the economic analyses that are undertaken.

Safety assessment of botanicals and botanical preparations used as ingredients in food supplements: testing an European Food Safety Authority-tiered approach.

Science.gov (United States)

Speijers, Gerrit; Bottex, Bernard; Dusemund, Birgit; Lugasi, Andrea; Tóth, Jaroslav; Amberg-Müller, Judith; Galli, Corrado L; Silano, Vittorio; Rietjens, Ivonne M C M

2010-02-01

This article describes results obtained by testing the European Food Safety Authority-tiered guidance approach for safety assessment of botanicals and botanical preparations intended for use in food supplements. Main conclusions emerging are as follows. (i) Botanical ingredients must be identified by their scientific (binomial) name, in most cases down to the subspecies level or lower. (ii) Adequate characterization and description of the botanical parts and preparation methodology used is needed. Safety of a botanical ingredient cannot be assumed only relying on the long-term safe use of other preparations of the same botanical. (iii) Because of possible adulterations, misclassifications, replacements or falsifications, and restorations, establishment of adequate quality control is necessary. (iv) The strength of the evidence underlying concerns over a botanical ingredient should be included in the safety assessment. (v) The matrix effect should be taken into account in the safety assessment on a case-by-case basis. (vi) Adequate data and methods for appropriate exposure assessment are often missing. (vii) Safety regulations concerning toxic contaminants have to be complied with. The application of the guidance approach can result in the conclusion that safety can be presumed, that the botanical ingredient is of safety concern, or that further data are needed to assess safety.

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)

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

Tensit - a novel probabilistic simulation tool for safety assessments. Tests and verifications using biosphere models

Energy Technology Data Exchange (ETDEWEB)

Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

2004-06-01

This report documents the verification of a new simulation tool for dose assessment put together in a package under the name Tensit (Technical Nuclide Simulation Tool). The tool is developed to solve differential equation systems describing transport and decay of radionuclides. It is capable of handling both deterministic and probabilistic simulations. The verifications undertaken shows good results. Exceptions exist only where the reference results are unclear. Tensit utilise and connects two separate commercial softwares. The equation solving capability is derived from the Matlab/Simulink software environment to which Tensit adds a library of interconnectable building blocks. Probabilistic simulations are provided through a statistical software named at{sub R}isk that communicates with Matlab/Simulink. More information about these softwares can be found at www.palisade.com and www.mathworks.com. The underlying intention of developing this new tool has been to make available a cost efficient and easy to use means for advanced dose assessment simulations. The mentioned benefits are gained both through the graphical user interface provided by Simulink and at{sub R}isk, and the use of numerical equation solving routines in Matlab. To verify Tensit's numerical correctness, an implementation was done of the biosphere modules for dose assessments used in the earlier safety assessment project SR 97. Acquired probabilistic results for deterministic as well as probabilistic simulations have been compared with documented values. Additional verification has been made both with another simulation tool named AMBER and also against the international test case from PSACOIN named Level 1B. This report documents the models used for verification with equations and parameter values so that the results can be recreated. For a background and a more detailed description of the underlying processes in the models, the reader is referred to the original references. Finally, in the

Tensit - a novel probabilistic simulation tool for safety assessments. Tests and verifications using biosphere models

International Nuclear Information System (INIS)

Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

2004-06-01

This report documents the verification of a new simulation tool for dose assessment put together in a package under the name Tensit (Technical Nuclide Simulation Tool). The tool is developed to solve differential equation systems describing transport and decay of radionuclides. It is capable of handling both deterministic and probabilistic simulations. The verifications undertaken shows good results. Exceptions exist only where the reference results are unclear. Tensit utilise and connects two separate commercial softwares. The equation solving capability is derived from the Matlab/Simulink software environment to which Tensit adds a library of interconnectable building blocks. Probabilistic simulations are provided through a statistical software named at R isk that communicates with Matlab/Simulink. More information about these softwares can be found at www.palisade.com and www.mathworks.com. The underlying intention of developing this new tool has been to make available a cost efficient and easy to use means for advanced dose assessment simulations. The mentioned benefits are gained both through the graphical user interface provided by Simulink and at R isk, and the use of numerical equation solving routines in Matlab. To verify Tensit's numerical correctness, an implementation was done of the biosphere modules for dose assessments used in the earlier safety assessment project SR 97. Acquired probabilistic results for deterministic as well as probabilistic simulations have been compared with documented values. Additional verification has been made both with another simulation tool named AMBER and also against the international test case from PSACOIN named Level 1B. This report documents the models used for verification with equations and parameter values so that the results can be recreated. For a background and a more detailed description of the underlying processes in the models, the reader is referred to the original references. Finally, in the perspective of

Assessment of 137Cs and 90Sr Fluxes in the Barents Sea

Science.gov (United States)

Matishov, Gennady; Usiagina, Irina; Kasatkina, Nadezhda; Ilin, Gennadii

2014-05-01

On the basis of published and own data the annual balance of radionuclide income/outcome was assessed for 137Cs and 90Sr in the Barents Sea for the period from 1950s to the presnt. The scheme of the isotope balance calculation in the Barents Sea included the following processes:atmospheric fallout; river run-off; liquid radioactive wastes releases, income from the Norwegian and the White Seas; outflow to the adjacent areas through the Novaya Zemlya straits and the transects Svalbard-Franz Josef Land and Franz Josef Land-Novaya Zemlya; radioactive decay. According to the multiyear dynamics, the inflow of 137Cs and 90Sr to the Barents Sea was significantly preconditioned by currents from the Norwegian Sea. Three peaks of 137Cs and 90Sr isotope concentrations were registered for the surface waters on the western border of the Barents Sea. The first one was observed in the mid-1960s and was conditioned by testing of nuclear weapons. The increase of isotope concentrations in 1975 and 1980 was preconditioned by the discharge of atomic waste by the Sellafield nuclear reprocessing plant. Nowadays, after the sewage disposal plant was built, the annual discharge of nuclear waste from Sellafield plant is low. The Norwegian Sea was a major source of 137Cs and 90Sr isotope income into the Barents Sea for the period of 1960-2014. Currently, the transborder transfer of 90Sr and 137Cs from the Norwegian Sea into the Barents Sea constitutes about 99% of income for each element. Atmospheric precipitation had a major impact in the 1950-1960s after the testing of the nuclear weapons, and in 1986 after the accident at Chernobyl Nuclear Power Station. In 1963, the atmospheric precipitation of 137Cs reached 1050 TBq; and that of 90Sr, 630 TBq. In 1986, a significant amount of 137Cs inflow (up to 1010 TBq/year) was registered. The 137Cs isotope income exceeded the 90Sr income in the 1960s-1980s, and equal amounts penetrated into the Barents Sea from the Norwegian Sea in the 1990s. Before

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

Formal safety assessment based on relative risks model in ship navigation

Energy Technology Data Exchange (ETDEWEB)

Hu Shenping [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: sphu@mmc.shmtu.edu.cn; Fang Quangen [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: qgfang@mmc.shmtu.edu.cn; Xia Haibo [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: hbxia@mmc.shmtu.edu.cn; Xi Yongtao [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: xiyt@mmc.shmtu.edu.cn

2007-03-15

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice.

Formal safety assessment based on relative risks model in ship navigation

International Nuclear Information System (INIS)

Hu Shenping; Fang Quangen; Xia Haibo; Xi Yongtao

2007-01-01

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice

Disposal of radioactive waste: can long-term safety be evaluated

International Nuclear Information System (INIS)

1991-01-01

The long-term safety of any hazardous waste disposal system must be convincingly shown prior to its implementation. For radioactive wastes, safety assessments over timescales far beyond the normal horizon of social and technical planning have already been conducted in many countries. These assessments provide the principal means to investigate, quantify, and explain long-term safety of each selected disposal concept and site for the appropriate authorities and the public. Such assessments are based on four main elements: definition of the disposal system and its environment, identification of possible processes and events that may affect the integrity of the disposal system, quantification of the radiological impact by predictive modelling, and description of associated uncertainties. The NEA Radioactive Waste Management Committee and the IAEA International Radioactive Waste Management Advisory Committee have carefully examined the current scientific methods for safety assessments of radioactive waste disposal systems, as briefly summarized in this report. The Committees have also reviewed the experience now available from using safety assessment methods in many countries, for different disposal concepts and formations, and in the framework of both nationally and internationally conducted studies, as referenced in this report [fr

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

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

Assessing progress in the development of safety culture

International Nuclear Information System (INIS)

Rotaru, I.; Ghita, S.; Biro, L.

2002-01-01

This paper is focussed on the organizational culture and learning processes required for the implementation of all aspects of safety culture. There is no prescriptive formula for improving safety culture. However, some common characteristics and practices are emerging that can be adopted by organizations in order to make progress. The paper refers to some approaches that have been successful in a number of countries. The experience of the international nuclear industry in the development and improvement of safety culture could be extended and found useful in other nuclear activities, irrespective of scale. The examples given of specific practice cover a wide range of activities including analysis of events, the regulatory approach on safety culture, employee participation and safety performance measures. Many of these practices may be relevant to smaller organizations and could contribute to improving safety culture, whatever the size of the organization. The most effective approach is to pursue a range of practices that can be mutually supportive in the development of a progressive safety culture, supported by professional standards, organizational and management commitment. Some guidance is also given on the assessment of safety culture and on the detection of a weakening safety culture. Few suggestions for accelerating the safety culture development and improvement process are also provided. (author)

Comments on geochemical aspects of SR 97

International Nuclear Information System (INIS)

Arthur, R.C.; Wei Zhou

2000-01-01

The Swedish Government has asked SKB to carry out a safety assessment of the KBS-3 disposal concept for spent nuclear fuel 'to demonstrate that the KBS-3 method has good prospects of being able to meet the safety and radiation protection requirements which SKI and SSI have specified in recent years.' The results of that assessment, referred to as SR 97, have recently been published. The present report summarizes the results of a review of selected geochemical aspects of SR 97. These subjects include the hydrochemical evolution of a defective canister, thermodynamic data supporting estimates of radioelement solubilities, modeling of near-field chemistry and analyses of the effects of ice melting on propagation of an oxidizing front to repository depths. The primary focus of the review is on the canister-defect scenario, and, more specifically, on supporting analyses of the hydromechanical evolution of a defective canister. The results of these analyses figure prominently in the safety assessment because they suggest that even a defective canister will, in effect, remain dry for as long as 200,000 years. This is an important constraint because it is taken in SR 97 as the period of time required for a continuous water pathway to form in the near field. The transport of most radionuclides (i.e., those that do not exist as a gas) cannot occur until this pathway is formed. It is concluded that although SKBs hydromechanical models are sound, they may suffer from an over-simplification of the chemical processes involved. Analyses using the models do not acknowledge that the chemical system within the canister is open in all respects to the chemical system in the buffer. Instead, mass transfer across the defect at the canister-buffer interface is limited to liquid H 2 O and water vapor. Consideration of mass transfer of other gases [e.g., CO 2 and H 2 S] dissolved in buffer porewaters suggests that associated reactions involving the iron insert and inner surfaces of the

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.

Deep repository for spent nuclear fuel. SR 97 - Post-closure safety. Main Report. Summary

International Nuclear Information System (INIS)

Hedin, A.

1999-11-01

structure of the geosphere and for earthquake statistics. The analysis method is new and includes several highly pessimistic simplifications. The analyses show that the probability of canister damage is comparable with the probability assumed for initial damage in the canister defect scenario. In the evaluation of the analysis method, it is shown how less pessimistic assumptions should lead to no canister damage at all in the model studies. The method will be refined. The scenario that deals with future inadvertent human actions that could conceivably affect the repository is surrounded by great uncertainties, chiefly because the evolution of human society is unpredictable. SR 97 discusses how conceivable societal evolutions and future human actions that affect the repository can nevertheless be categorized to some extent. In an illustrative example, a situation is analyzed where a canister in the repository is inadvertently penetrated by rock drillers. Dose and risk are calculated for the drilling personnel and for a family that settles on the site at a later time. The principal conclusion of the SR 97 safety assessment is that the prospects of building a safety deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a programme for site investigations, for formulating functional requirements on the repository's barriers, and for prioritization of research. The next stage in the siting of a deep repository entails investigation of the bedrock at a number of candidate sites in Sweden. It is SKBs judgement that the scope of the safety assessment and confidence in its results satisfy the requirements that should be made in preparation for such a stage

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

Effect of Wood Aging on Wine Mineral Composition and 87Sr/86Sr Isotopic Ratio.

Science.gov (United States)

Kaya, Ayse D; Bruno de Sousa, Raúl; Curvelo-Garcia, António S; Ricardo-da-Silva, Jorge M; Catarino, Sofia

2017-06-14

The evolution of mineral composition and wine strontium isotopic ratio 87 Sr/ 86 Sr (Sr IR) during wood aging were investigated. A red wine was aged in stainless steel tanks with French oak staves (Quercus sessiliflora Salisb.), with three industrial scale replicates. Sampling was carried out after 30, 60, and 90 days of aging, and the wines were evaluated in terms of general analysis, phenolic composition, total polysaccharides, multielement composition, and Sr IR. Li, Be, Mg, Al, Sc, Ti, V, Mn, Co, Ni, Cu, Zn, Ga, Ge, As, Rb, Sr, Y, Zr, Mo, Sb, Cs, Ba, Pr, Nd, Sm, Eu, Dy, Ho, Er, Yb, Lu, Tl, and Pb elements and 87 Sr/ 86 Sr were determined by quadrupole inductively coupled plasma mass spectrometry (Q-ICP-MS) and Na, K, Ca, and Fe by flame atomic absorption spectrometry (FAAS). Two-way ANOVA was applied to assess wood aging and time effect on Sr IR and mineral composition. Wood aging resulted in significantly higher concentrations of Mg, V, Co, Ni, and Sr. At the end of the aging period, wine exhibited statistically identical Sr IR compared to control. Study suggests that wood aging does not affect 87 Sr/ 86 Sr, not precluding the use of this parameter for wine traceability purposes.

Selected component failure rate values from fusion safety assessment tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee Charles

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected component failure rate values from fusion safety assessment tasks

International Nuclear Information System (INIS)

Cadwallader, L.C.

1998-01-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers

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

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.

Strontium (Sr) separation from seawater using titanate adsorbents: Effects of seawater matrix ions on Sr sorption behavior

Science.gov (United States)

Ryu, Jungho; Hong, Hye-jin; Ryu, Taegong; Park, In-Su

2017-04-01

Strontium (Sr) which has many industrial applications such as ferrite magnet, ceramic, and fire works exists in seawater with the concentration of approximately 7 mg/L. In previous report estimating economic potential on recovery of various elements from seawater in terms of their commercial values and concentrations in seawater, Sr locates upper than approximate break-even line, which implies Sr recovery from seawater can be potentially profitable. Recently, Sr separation from seawater has received great attention in the environmental aspect after Fukushima Nuclear Power Plant (NPP) accident which released much amount of radioactive Sr and Cs. Accordingly, the efficient separation of radioactive elements released to seawater has become critical as an important technological need as well as their removal from radioactive wastes. So far, it has been introduced to separate Sr from aqueous media by various methods including solvent extraction, adsorption by solid materials, and ion exchange. Among them, the adsorption technique using solid adsorbents is of great interest for selectively separating Sr from seawater with respect to low concentration level of Sr. In this study, we synthesized titanate nanotube (TiNT) by simple hydrothermal reaction, characterized its physicochemical properties, and systematically evaluated Sr sorption behavior under various reaction conditions corresponding to seawater environment. The synthesized TiNT exhibited the fibril-type nanotube structure with high specific surface area of 260 m2/g. The adsorption of Sr on TiNT rapidly occurred following pseudo-second-order kinetic model, and was in good agreement with Langmuir isotherm model, indicating maximum adsorption capacity of 97 mg/g. Based on Sr uptake and Na release with stoichiometric balance, sorption mechanism of Sr on TiNT was found to be ion-exchange between Na in TiNT lattice and Sr in solution phase, which was also confirmed by XRD and Raman analysis. Among competitive ions, Ca

Can cyclist safety be improved with intelligent transport systems?

Science.gov (United States)

Silla, Anne; Leden, Lars; Rämä, Pirkko; Scholliers, Johan; Van Noort, Martijn; Bell, Daniel

2017-08-01

In recent years, Intelligent Transport Systems (ITS) have assisted in the decrease of road traffic fatalities, particularly amongst passenger car occupants. Vulnerable Road Users (VRUs) such as pedestrians, cyclists, moped riders and motorcyclists, however, have not been that much in focus when developing ITS. Therefore, there is a clear need for ITS which specifically address VRUs as an integrated element of the traffic system. This paper presents the results of a quantitative safety impact assessment of five systems that were estimated to have high potential to improve the safety of cyclists, namely: Blind Spot Detection (BSD), Bicycle to Vehicle communication (B2V), Intersection safety (INS), Pedestrian and Cyclist Detection System+Emergency Braking (PCDS+EBR) and VRU Beacon System (VBS). An ex-ante assessment method proposed by Kulmala (2010) targeted to assess the effects of ITS for cars was applied and further developed in this study to assess the safety impacts of ITS specifically designed for VRUs. The main results of the assessment showed that all investigated systems affect cyclist safety in a positive way by preventing fatalities and injuries. The estimates considering 2012 accident data and full penetration showed that the highest effects could be obtained by the implementation of PCDS+EBR and B2V, whereas VBS had the lowest effect. The estimated yearly reduction in cyclist fatalities in the EU-28 varied between 77 and 286 per system. A forecast for 2030, taking into accounts the estimated accident trends and penetration rates, showed the highest effects for PCDS+EBR and BSD. Copyright © 2016 Elsevier Ltd. All rights reserved.

Plasma-safety assessment model and safety analyses of ITER

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

2001-01-01

A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

International cooperation in the safety and environmental assessment for the ITER engineering design activities

International Nuclear Information System (INIS)

Gordon, C.; Baker, D.J.; Bartels, H-W.

1998-01-01

The ITER Project includes design and assessment activities to ensure the safety and environmental attractiveness of ITER and demonstrate that it can be sited in any of the sponsoring Parties with a minimum of site-specific redesign. This paper highlights some of the efforts to develop an international consensus approach for ITER safety design and assessment, including: development of general safety and environmental design criteria; development of quantitative dose-release assessment criteria; development of a radiation protection program; waste characterization; and development of safety analysis guidelines. The high level of interaction, cooperation and collaboration between the Joint Central Team and the Home Teams, and between the safety team and designers, and the spirit of consensus that has guided them have resulted in a safe design for ITER and a safety design and assessment that can meet the needs of the potential host countries. (author)

Thermodynamic modeling of the Sr-Co-Fe-O system

DEFF Research Database (Denmark)

Zhang, Wei Wei; Povoden-Karadeniz, Erwin; Chen, Ming

2016-01-01

This paper reviews and assesses phase equilibria and thermodynamic properties of phases in the Sr-Co-Fe-O system, with a focus on oxides, especially the SrCo1 - xFexO3 - δ perovskite. In our work, the SrCo1 - xFexO3 - δ perovskite was modeled with a three-sublattice model, where the three...... sublattices correspond to the A, B and oxygen sites in an ABO3 perovskite, respectively. A number of other important ternary oxide phases in Sr-Co-O and Sr-Co-Fe-O were also considered. Available thermodynamic and phase diagram data were carefully assessed. A thermodynamic description of Sr-Co-O was derived...

Safety assessment of multi-unit NPP sites subject to external events

International Nuclear Information System (INIS)

Samaddar, Sujit; Hibino, Kenta; Coman, Ovidiu

2014-01-01

This paper presents a framework for conducting a probabilistic safety assessment of multi-unit sites against external events. The treatment of multiple hazard on a unit, interaction between units, implementation of severe accident measures, human reliability, environmental conditions, metric of risk for both reactor and non-reactor sources, integration of risk and responses and many such important factors need to be addressed within the context of this framework. The framework facilitates the establishment of a comprehensive methodology that can be applied internationally to the peer review of safety assessment of multi-unit sites under the impact of multiple external hazards. In summary, it can be said that the site safety assessment for a multi-unit site will be quite complex and need to start with individual unit risk assessments, these need to be combined considering the interactions between units and their responses, and the fragilities of the installations established considering the combined demands from all interactions. Using newly established risk metric the risk can then be integrated for the overall site. Fig. 2 shows schematically such a proposal. Much work has to done and the IAEA has established a working group that is systematically establishing the structure and process to incorporate the many issues that are a part of a multi-unit site safety assessment. (authors)

Assessment of the Safety of Some On-The-Shelf Canned Food ...

African Journals Online (AJOL)

There is also the possibility of these organisms posing food safety issues and pharmaceutical risks in case of possible out break, assayed through plasmid profiling of the culture-dependent isolates. A major concern in this study is the lack of adherence to food safety regulations. The products still been marketed on the ...

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

Method of operator safety assessment for underground mobile mining equipment

Science.gov (United States)

Działak, Paulina; Karliński, Jacek; Rusiński, Eugeniusz

2018-01-01

The paper presents a method of assessing the safety of operators of mobile mining equipment (MME), which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM). As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

Method of operator safety assessment for underground mobile mining equipment

Directory of Open Access Journals (Sweden)

Działak Paulina

2018-01-01

Full Text Available The paper presents a method of assessing the safety of operators of mobile mining equipment (MME, which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM. As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

Safety of nuclear installations in the Slovak Republic and activities of the Nuclear Regulatory Authority of the Slovak Republic in 2007

International Nuclear Information System (INIS)

Zemanova, D.

2008-01-01

Prepared pursuant to the provisions of the Atomic Act, the report provides information on the safety of nuclear installation in the Slovak Republic and activities of the Nuclear Regulatory Authority of the Slovak Republic ( UJD SR). UJD SR executes its activities in the area of legislation, issuance of authorizations and permissions for the siting, construction, operation and decommissioning of nuclear installations, in the area of reviews, assessments and control of nuclear safety of nuclear installations and emergency planning, in the area of records and accountability of nuclear materials, independent public information and in the area of international co-operation focused on peaceful uses of nuclear power. Based on the results of inspection activities and evaluation of safety indicators, UJD SR assessed the operation of nuclear installations in the Slovak Republic as safe and reliable. No significant event that could have a negative impact on the personnel, population or environment occurred in 2007. (orig.)

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

Isotope ratio 87Sr/86Sr in limestones from Bambui group, Brazil (MG)

International Nuclear Information System (INIS)

Kawashita, K.; Mizusaki, A.M.P.; Kiang, C.H.

1987-01-01

The Sr composition of ancient seawater can be estimated from the analysis of carbonate rocks and, in some cases, used to estimate the age of the analyzed carbonate. The normalized 87Sr/86Sr ratios in calcium carbonate fractions from 14 core samples in the Bambui Group near Montalvania, MG, were found to range between .7077 and .7280. The higher values are attributable to Sr isotopic exchange between silicate and carbonate phases during diagenesis. The ratio of .7077 obtained in two pure calcium carbonate samples is here suggested as the best aproximation for the 87Sr/86Sr value for the Bambui sea. This ratio is compatible with an age of about 700 Ma., estimated from the published 87Sr/86Sr curve of Veizer and others, an age in accordance with Quadros recent (1987, in preparation) identification of marine acritarchs from the latest Precambrian (Vendian). (author) [pt

A new approach to assess the effects of Sr and Bi interaction in ADC12 Al–Si die casting alloy

International Nuclear Information System (INIS)

Farahany, Saeed; Ourdjini, Ali; Abu Bakar, Tuty Asma; Idris, Mohd Hasbullah

2014-01-01

Highlights: • Interactive effect between Bi and Sr has been invesitigated comprehensively. • Sequence of addition did not affect thermal and microscopical characteristics. • A new map has been established to assess the final microstructure of castings. - Abstract: In the present paper, the possible interaction between bismuth and strontium in ADC12 die casting alloy was investigated comprehensively by using in situ thermal analysis technique. The characteristic temperatures including nucleation, minimum and growth temperatures of eutectic Al–Si were also analyzed. The results show that with Bi present in the Al–Si alloy melt the efficiency of Sr in modifying the eutectic Si is reduced. A threshold Sr/Bi ratio of at least 0.5 is required for a fully modified Si structure to form. A new map based on the characteristic temperatures, Sr/Bi ratio and microstructure, was established to assess the microstructure of fully solidified Al–Si castings

A new approach to assess the effects of Sr and Bi interaction in ADC12 Al–Si die casting alloy

Energy Technology Data Exchange (ETDEWEB)

Farahany, Saeed, E-mail: saeedfarahany@gmail.com; Ourdjini, Ali; Abu Bakar, Tuty Asma; Idris, Mohd Hasbullah

2014-01-10

Highlights: • Interactive effect between Bi and Sr has been invesitigated comprehensively. • Sequence of addition did not affect thermal and microscopical characteristics. • A new map has been established to assess the final microstructure of castings. - Abstract: In the present paper, the possible interaction between bismuth and strontium in ADC12 die casting alloy was investigated comprehensively by using in situ thermal analysis technique. The characteristic temperatures including nucleation, minimum and growth temperatures of eutectic Al–Si were also analyzed. The results show that with Bi present in the Al–Si alloy melt the efficiency of Sr in modifying the eutectic Si is reduced. A threshold Sr/Bi ratio of at least 0.5 is required for a fully modified Si structure to form. A new map based on the characteristic temperatures, Sr/Bi ratio and microstructure, was established to assess the microstructure of fully solidified Al–Si castings.

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

Accessible switching of electronic defect type in SrTi O3 via biaxial strain

Science.gov (United States)

Chi, Yen-Ting; Youssef, Mostafa; Sun, Lixin; Van Vliet, Krystyn J.; Yildiz, Bilge

2018-05-01

Elastic strain is used widely to alter the mobility of free electronic carriers in semiconductors, but a predictive relationship between elastic lattice strain and the extent of charge localization of electronic defects is still underdeveloped. Here we considered SrTi O3 , a prototypical perovskite as a model functional oxide for thin film electronic devices and nonvolatile memories. We assessed the effects of biaxial strain on the stability of electronic defects at finite temperature by combining density functional theory (DFT) and quasiharmonic approximation (QHA) calculations. We constructed a predominance diagram for free electrons and small electron polarons in this material, as a function of biaxial strain and temperature. We found that biaxial tensile strain in SrTi O3 can stabilize the small polaron, leading to a thermally activated and slower electronic transport, consistent with prior experimental observations on SrTi O3 and distinct from our prior theoretical assessment of the response of SrTi O3 to hydrostatic stress. These findings also resolved apparent conflicts between prior atomistic simulations and conductivity experiments for biaxially strained SrTi O3 thin films. Our computational approach can be extended to other functional oxides, and for the case of SrTi O3 our findings provide concrete guidance for conditions under which strain engineering can shift the electronic defect type and concentration to modulate electronic transport in thin films.

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

Establishment of safety goal and its quantification based on risk assessment

International Nuclear Information System (INIS)

Miyano, Hiroshi; Muramatsu, Ken

2017-01-01

We must clarify the safety objectives sought by society in securing the safety of nuclear reactors and nuclear power plants. For that purpose, it is useful to utilize risk assessment. Quantitative methods including probabilistic risk assessment (PRA) are superior in terms of scientific rationality and quantitative performance compared with conventional deterministic methods, and able to indicate an objective numerical value of safety level. Consequently, quantitative methods can enhance the transparency, consistency, compliance, predictability, and explanatory power of regulatory decisions toward business operators and citizens. Business operators can explain the validity of their own safety assurance activities to regulators and citizens. The goal to be secured becomes clear by incorporating the safety goal into the specific performance goal required for the nuclear power plant from the viewpoint of deep safeguard, and it becomes easy to evaluate the effectiveness of the safety measures. It helps us greatly in judging and selecting the appropriateness of safety measures. It should be noted: the fact that the result of implementing the PRA satisfies the safety goal is not a sufficient condition in the sense of guaranteeing complete safety but a necessary condition. The nuclear power field is a region with large uncertainty, and research/efforts for accuracy improvement and evaluation validity will be required continuously. (A.O.)

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)

Structural and compositional characterization of synthetic (Ca,Sr)-tremolite and (Ca,Sr)-diopside solid solutions

Science.gov (United States)

Gottschalk, M.; Najorka, J.; Andrut, M.

Tremolite (CaxSr1-x)2Mg5[Si8O22/(OH)2] and diopside (CaxSr1-x)Mg[Si2O6] solid solutions have been synthesized hydrothermally in equilibrium with a 1 molar (Ca,Sr)Cl2 aqueous solution at 750°C and 200 MPa. The solid run products have been investigated by optical, electron scanning and high resolution transmission electron microscopy, electron microprobe, X-ray-powder diffraction and Fourier-transform infrared spectroscopy. The synthesized (Ca,Sr)-tremolites are up to 2000 µm long and 30 µm wide, the (Ca,Sr)-diopsides are up to 150 µm long and 20 µm wide. In most runs the tremolites and diopsides are well ordered and chain multiplicity faults are rare. Nearly pure Sr-tremolite (tr0.02Sr-tr0.98) and Sr-diopside (di0.01Sr-di0.99) have been synthesized. A continuous solid solution series, i.e. complete substitution of Sr2+ for Ca2+ on M4-sites exists for (Ca,Sr)-tremolite. Total substitution of Sr2+ for Ca2+ on M2-sites can be assumed for (Ca,Sr)-diopsides. For (Ca,Sr)-tremolites the lattice parameters a, b and β are linear functions of composition and increase with Sr-content whereas c is constant. For the diopside series all 4 lattice parameters are a linear function of composition; a, b, c increase and β decreases with rising Sr-content. The unit cell volume for tremolite increases 3.47% from 906.68 Å3 for tremolite to 938.21 Å3 for Sr-tremolite. For diopside the unit cell volume increases 4.87 % from 439.91 Å3 for diopside to 461.30 Å3 for Sr-diopside. The observed splitting of the OH stretching band in tremolite is caused by different configurations of the next nearest neighbors (multi mode behavior). Resolved single bands can be attributed to the following configurations on the M4-sites: SrSr, SrCa, CaCa and CaMg. The peak positions of these 4 absorption bands are a linear function of composition. They are shifted to lower wavenumbers with increasing Sr-content. No absorption band due to the SrMg configuration on the M4-site is observed. This indicates

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

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)

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

2005-03-01

Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

Quantitative reliability assessment for safety critical system software

International Nuclear Information System (INIS)

Chung, Dae Won; Kwon, Soon Man

2005-01-01

An essential issue in the replacement of the old analogue I and C to computer-based digital systems in nuclear power plants is the quantitative software reliability assessment. Software reliability models have been successfully applied to many industrial applications, but have the unfortunate drawback of requiring data from which one can formulate a model. Software which is developed for safety critical applications is frequently unable to produce such data for at least two reasons. First, the software is frequently one-of-a-kind, and second, it rarely fails. Safety critical software is normally expected to pass every unit test producing precious little failure data. The basic premise of the rare events approach is that well-tested software does not fail under normal routine and input signals, which means that failures must be triggered by unusual input data and computer states. The failure data found under the reasonable testing cases and testing time for these conditions should be considered for the quantitative reliability assessment. We will present the quantitative reliability assessment methodology of safety critical software for rare failure cases in this paper

Thermodynamic modeling of the Sr-Co-Fe-O system

DEFF Research Database (Denmark)

Zhang, Wei Wei; Povoden-Karadeniz, Erwin; Chen, Ming

2016-01-01

This paper reviews and assesses phase equilibria and thermodynamic properties of phases in the Sr-Co-Fe-O system, with a focus on oxides, especially the SrCo1 - xFexO3 - δ perovskite. In our work, the SrCo1 - xFexO3 - δ perovskite was modeled with a three-sublattice model, where the three...... sublattices correspond to the A, B and oxygen sites in an ABO3 perovskite, respectively. A number of other important ternary oxide phases in Sr-Co-O and Sr-Co-Fe-O were also considered. Available thermodynamic and phase diagram data were carefully assessed. A thermodynamic description of Sr-Co-O was derived...... using the CALPHAD approach and was further extrapolated to that of Sr-Co-Fe-O. The thermodynamic database of Sr-Co-Fe-O established in this work allows for calculating phase diagrams, thermodynamic properties, cation distribution and defect chemistry properties, and therefore enables material...

Operational safety assessment of underground test facilities for mined geologic waste disposal

International Nuclear Information System (INIS)

Elder, H.K.

1993-01-01

This paper describes the operational safety assessment for the underground facilities for the exploratory studies facility (ESF) at the Yucca Mountain Project. The systematic identification and evaluation of hazards related to the ESF is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach based on the analysis of potential accidents was used since radiological safety analysis was not required. The risk assessment summarized credible accident scenarios and the design provides mitigation of the risks to a level that the facility can be constructed and operated with an adequate level of safety. The risk assessment also provides reasonable assurance that all identifiable major accident scenarios have been reviewed and design mitigation features provided to ensure an adequate level of safety

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

Crane Safety Assessment Method Based on Entropy and Cumulative Prospect Theory

Directory of Open Access Journals (Sweden)

Aihua Li

2017-01-01

Full Text Available Assessing the safety status of cranes is an important problem. To overcome the inaccuracies and misjudgments in such assessments, this work describes a safety assessment method for cranes that combines entropy and cumulative prospect theory. Firstly, the proposed method transforms the set of evaluation indices into an evaluation vector. Secondly, a decision matrix is then constructed from the evaluation vectors and evaluation standards, and an entropy-based technique is applied to calculate the index weights. Thirdly, positive and negative prospect value matrices are established from reference points based on the positive and negative ideal solutions. Thus, this enables the crane safety grade to be determined according to the ranked comprehensive prospect values. Finally, the safety status of four general overhead traveling crane samples is evaluated to verify the rationality and feasibility of the proposed method. The results demonstrate that the method described in this paper can precisely and reasonably reflect the safety status of a crane.

Implementing systematic review techniques in chemical risk assessment: Challenges, opportunities and recommendations.

Science.gov (United States)

Whaley, Paul; Halsall, Crispin; Ågerstrand, Marlene; Aiassa, Elisa; Benford, Diane; Bilotta, Gary; Coggon, David; Collins, Chris; Dempsey, Ciara; Duarte-Davidson, Raquel; FitzGerald, Rex; Galay-Burgos, Malyka; Gee, David; Hoffmann, Sebastian; Lam, Juleen; Lasserson, Toby; Levy, Len; Lipworth, Steven; Ross, Sarah Mackenzie; Martin, Olwenn; Meads, Catherine; Meyer-Baron, Monika; Miller, James; Pease, Camilla; Rooney, Andrew; Sapiets, Alison; Stewart, Gavin; Taylor, David

2016-01-01

Systematic review (SR) is a rigorous, protocol-driven approach designed to minimise error and bias when summarising the body of research evidence relevant to a specific scientific question. Taking as a comparator the use of SR in synthesising research in healthcare, we argue that SR methods could also pave the way for a "step change" in the transparency, objectivity and communication of chemical risk assessments (CRA) in Europe and elsewhere. We suggest that current controversies around the safety of certain chemicals are partly due to limitations in current CRA procedures which have contributed to ambiguity about the health risks posed by these substances. We present an overview of how SR methods can be applied to the assessment of risks from chemicals, and indicate how challenges in adapting SR methods from healthcare research to the CRA context might be overcome. Regarding the latter, we report the outcomes from a workshop exploring how to increase uptake of SR methods, attended by experts representing a wide range of fields related to chemical toxicology, risk analysis and SR. Priorities which were identified include: the conduct of CRA-focused prototype SRs; the development of a recognised standard of reporting and conduct for SRs in toxicology and CRA; and establishing a network to facilitate research, communication and training in SR methods. We see this paper as a milestone in the creation of a research climate that fosters communication between experts in CRA and SR and facilitates wider uptake of SR methods into CRA. Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.

Determination of mineral contents of wild Boletus edulis mushroom and its edible safety assessment.

Science.gov (United States)

Su, Jiuyan; Zhang, Ji; Li, Jieqing; Li, Tao; Liu, Honggao; Wang, Yuanzhong

2018-04-06

This study aimed to determine the contents of main mineral elements of wild Boletus edulis and to assess its edible safety, which may provide scientific evidence for the utilization of this species. Fourteen mineral contents (Ba, Ca, Cd, Co, Cr, Cu, Fe, Mg, Mn, Na, Ni, Sr, V and Zn) in the caps and stipes of B. edulis as well as the corresponding surface soils collected from nine different geographic regions in Yunnan Province, southwest China were determined. The analyses were performed using inductively coupled plasma atomic emission spectrometer (ICP-AES) after microwave digestion. Measurement data were analyzed using variance and Pearson correlation analysis. Edible safety was evaluated according to the provisionally tolerable weekly intake (PTWI) of heavy metals recommended by United Nations Food and Agriculture Organization and World Health Organization (FAO/WHO). Mineral contents were significantly different with the variance of collection areas. B. edulis showed relative abundant contents of Ca, Fe, Mg and Na, followed by Ba, Cr, Cu, Mn and Zn, and the elements with the lower content less were Cd, Co, Ni, Sr and V. The elements accumulation differed significantly in caps and stipes. Among them, Cd and Zn were bioconcentrated (BCF > 1) while others were bioexcluded (BCF < 1). The mineral contents in B. edulis and its surface soil were positively related, indicating that the elements accumulation level was related to soil background. In addition, from the perspective of food safety, if an adult (60 kg) eats 300 g fresh B. edulis per week, the intake of Cd in most of tested mushrooms were lower than PTWI value whereas the Cd intakes in some other samples were higher than this standard. The results indicated that the main mineral contents in B. edulis were significantly different with respect to geographical distribution, and the Cd intake in a few of regions was higher than the acceptable intakes with a potential risk.

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

Assessment of Sr-90 in water samples: precision and accuracy

International Nuclear Information System (INIS)

Nisti, Marcelo B.; Saueia, Cátia H.R.; Castilho, Bruna; Mazzilli, Barbara P.

2017-01-01

The study of artificial radionuclides dispersion into the environment is very important to control the nuclear waste discharges, nuclear accidents and nuclear weapons testing. The accidents in Fukushima Daiichi Nuclear Power Plant and Chernobyl Nuclear Power Plant, released several radionuclides in the environment by aerial deposition and liquid discharge, with various level of radioactivity. The 90 Sr was one of the elements released into the environment. The 90 Sr is produced by nuclear fission with a physical half-life of 28.79 years with decay energy of 0.546 MeV. The aims of this study are to evaluate the precision and accuracy of three methodologies for the determination of 90 Sr in water samples: Cerenkov, LSC direct method and with radiochemical separation. The performance of the methodologies was evaluated by using two scintillation counters (Quantulus and Hidex). The parameters Minimum Detectable Activity (MDA) and Figure Of Merit (FOM) were determined for each method, the precision and accuracy were checked using 90 Sr standard solutions. (author)

[Agricultural biotechnology safety assessment].

Science.gov (United States)

McClain, Scott; Jones, Wendelyn; He, Xiaoyun; Ladics, Gregory; Bartholomaeus, Andrew; Raybould, Alan; Lutter, Petra; Xu, Haibin; Wang, Xue

2015-01-01

Genetically modified (GM) crops were first introduced to farmers in 1995 with the intent to provide better crop yield and meet the increasing demand for food and feed. GM crops have evolved to include a thorough safety evaluation for their use in human food and animal feed. Safety considerations begin at the level of DNA whereby the inserted GM DNA is evaluated for its content, position and stability once placed into the crop genome. The safety of the proteins coded by the inserted DNA and potential effects on the crop are considered, and the purpose is to ensure that the transgenic novel proteins are safe from a toxicity, allergy, and environmental perspective. In addition, the grain that provides the processed food or animal feed is also tested to evaluate its nutritional content and identify unintended effects to the plant composition when warranted. To provide a platform for the safety assessment, the GM crop is compared to non-GM comparators in what is typically referred to as composition equivalence testing. New technologies, such as mass spectrometry and well-designed antibody-based methods, allow better analytical measurements of crop composition, including endogenous allergens. Many of the analytical methods and their intended uses are based on regulatory guidance documents, some of which are outlined in globally recognized documents such as Codex Alimentarius. In certain cases, animal models are recommended by some regulatory agencies in specific countries, but there is typically no hypothesis or justification of their use in testing the safety of GM crops. The quality and standardization of testing methods can be supported, in some cases, by employing good laboratory practices (GLP) and is recognized in China as important to ensure quality data. Although the number of recommended, in some cases, required methods for safety testing are increasing in some regulatory agencies, it should be noted that GM crops registered to date have been shown to be

Role and meaning of safety assessment from the point of view of IAEA

International Nuclear Information System (INIS)

Lyubarskiy, A.

2012-01-01

In 2006, the IAEA published its revised Safety Fundamentals. This states that the ''fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation''. This objective has to be achieved for all facilities and activities and for all stages over the lifetime of a facility by adherence to ten fundamental principles. This leads, inter alia, to the requirement for a safety assessment to be carried out. In particular, the text accompanying Principle 3 on leadership and management for safety states that: ''3.15. Safety has to be assessed for all facilities and activities, consistent with a graded approach. Safety assessment involves the systematic analysis of normal operation and its effects, of the ways in which failures might occur and of the consequences of such failures. Safety assessments cover the safety measures necessary to control the hazard, and the design and engineered safety features are assessed to demonstrate that they fulfill the safety functions required of them. Where control measures or operator actions are called on to maintain safety, an initial safety assessment has to be carried out to demonstrate that the arrangements made are robust and that they can be relied on. A facility may only be constructed and commissioned or an activity may only be commenced once it has been demonstrated to the satisfaction of the regulatory body that the proposed safety measures are adequate.'' Principle 3 further states that the process of safety assessment for facilities and activities is repeated in the conduct of operations in order to take into account changed circumstances (such as the application of new standards or scientific and technological developments), the feedback of operating experience, modifications and the effects of ageing. Continuation of operations over long periods of time requires reassessments demonstrating that the safety measures remain adequate. (orig.)

Validity of instruments to assess students' travel and pedestrian safety

Directory of Open Access Journals (Sweden)

Baranowski Tom

2010-05-01

Full Text Available Abstract Background Safe Routes to School (SRTS programs are designed to make walking and bicycling to school safe and accessible for children. Despite their growing popularity, few validated measures exist for assessing important outcomes such as type of student transport or pedestrian safety behaviors. This research validated the SRTS school travel survey and a pedestrian safety behavior checklist. Methods Fourth grade students completed a brief written survey on how they got to school that day with set responses. Test-retest reliability was obtained 3-4 hours apart. Convergent validity of the SRTS travel survey was assessed by comparison to parents' report. For the measure of pedestrian safety behavior, 10 research assistants observed 29 students at a school intersection for completion of 8 selected pedestrian safety behaviors. Reliability was determined in two ways: correlations between the research assistants' ratings to that of the Principal Investigator (PI and intraclass correlations (ICC across research assistant ratings. Results The SRTS travel survey had high test-retest reliability (κ = 0.97, n = 96, p Conclusions These validated instruments can be used to assess SRTS programs. The pedestrian safety behavior checklist may benefit from further formative work.

You can't improve what you don't measure: Safety climate measures available in the German-speaking countries to support safety culture development in healthcare.

Science.gov (United States)

Manser, Tanja; Brösterhaus, Mareen; Hammer, Antje

2016-01-01

Safety climate measurement is a key input into safety culture development. The aim of this review is to provide an overview of the safety climate measures that have been evaluated for their psychometric properties in a German-speaking country and to make recommendations on how to use them in quality and patient safety improvement. A systematic search strategy was implemented to obtain relevant articles. PubMed and Web of Science databases were searched, and 128 abstracts were identified. After application of limits, 33 full texts were retrieved for subsequent evaluation. Studies were included on the basis of predetermined inclusion criteria and independent assessment by two reviewers. Publications were reviewed concerning healthcare setting, target group, safety culture dimensions covered and results of their psychometric evaluation. This review identified 11 instruments for safety climate assessment in different healthcare settings (i. e. hospitals, nursing homes, primary care, dental care and community pharmacy) for which acceptable to good internal consistency was reported. We observed wide variability concerning the number of dimensions (1 to 14; in some cases including outcome dimensions) and items (9 to 128) that the instruments were comprised of. Nevertheless, consistency with regard to the thematic areas covered was rather high. While there is clear evidence that we can assess safety climate in healthcare, the application of safety climate measures by quality and patient safety practitioners has so far been rather limited. This review bridges this gap between research and improvement practice by highlighting the central role of safety climate assessment in a mixed methods approach to inform safety culture development. Copyright © 2016. Published by Elsevier GmbH.

Safety assessment of a borehole type disposal facility using the ISAM methodology

International Nuclear Information System (INIS)

Blerk, J.J. van; Yucel, V.; Kozak, M.W.; Moore, B.A.

2002-01-01

As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to test the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Borehole Test Case (BTC), related to a proposed future disposal option for disused sealed radioactive sources. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the BTC and provides some general conclusions that can be drawn from the findings of their work. (author)

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)

Safety Assessment Context for Croatian Low and Intermediate Level Radioactive Waste Repository

International Nuclear Information System (INIS)

Levanat, I.; Lokner, V.

1998-01-01

Safety assessments in a small country are usually performed to support the national waste management strategy, demonstrating compliance with national regulation for a particular facility. However, this assessment should - quite generally - provide reasonable assurance both to the public and to decision makers than the Croatian share of LILW from NPP Krsko can be safely disposed in Croatia. More specifically, assessment should clearly present all realistic options and compare the associated long term repository performances, demonstrating that desirable safety goals can be archived by an appropriate choice of (a) location, (b) facility design, (c) institutional control period and (d) waste acceptance criteria. As relevant national legislation is presently under review, generally recognized international safety standards, criteria and recommendations (e.g. as presented in the recent IAEA publications) should provide guidance for the assessment evaluation, since it is expected that they will be incorporated in the new national regulations. Finally, since Croatian radioactive waste management strategy is yet to be developed, such an assessment may contribute to its formulation and facilitate some specific decisions. (author)

Towards understanding work-as-done in air traffic management safety assessment and design

International Nuclear Information System (INIS)

Woltjer, Rogier; Pinska-Chauvin, Ella; Laursen, Tom; Josefsson, Billy

2015-01-01

This paper describes the approach taken and the results to develop guidance, to include Resilience Engineering principles in methodology for safety assessment of functional changes, in Air Traffic Management (ATM). It summarizes the process of deriving resilience principles for ATM, originating from Resilience Engineering concepts and transposed into ATM operations. These principles are the foundation for guidance material incorporating Resilience Engineering (RE) concepts into safety assessment methodology. The guidance material provides a method using workshops generating qualitative descriptions of RE principles applied to ATM services of everyday work, as done currently and as envisioned after introduction of a new technology or way of working. The guidance material has been proposed as part of the safety assessment methodology of SESAR (Single European Sky ATM Research), and as stand-alone guidance for ATM design processes. The methodology was validated via a test case on the i4D/CTA (Controlled Time of Arrival) concept. Operational examples from the application of the developed guidance to the i4D/CTA concept are provided. Initial evaluation of the guidance suggests that the methodology (1) provides a narrative, vocabulary and documentation means of project discussions on resilience; (2) brings the discussions of safety and resilience closer to operational practice; (3) facilitates a broader systemic and integrative perspective on operational, management, business, safety, environmental, and human performance aspects; and (4) can extend the vocabulary of safety assessment to include the description of emergent properties, to better support functional changes in ATM. - Highlights: • Guidance material for safety assessment based on systemic thinking is proposed. • It operationalizes Resilience Engineering principles in Air Traffic Management, including a case study. • It enables description of expected changes in work-as-done when introducing a new

SR-Can. Data and uncertainty assessment. Matrix diffusivity and porosity in situ

International Nuclear Information System (INIS)

Jinsong Liu; Loefgren, Martin; Neretnieks, Ivars

2006-12-01

and show beyond doubt that the electrical conductivity method using AC gives the expected information on transport properties (diffusivities) of the pores of crystalline rocks. The electrical conductivity method is much faster and can be used to measure large samples. It has recently been adopted for use in deep boreholes. Tens of thousands of measurements have been made at Simpevarp, Laxemar and Forsmark (Swedish sites) at depths of more than 1,000 m. The results of these measurements form the basis for our proposed diffusion values to be used in Performance Assessment (PA) for the candidate sites (Performance Assessment (PA) used in this report is synonymous to Safety Assessment (SA) sometimes used by other authors). The in situ data are obtained essentially in undisturbed rock and have not been subject to either stress release or disturbances due to sample preparation. The small disturbance nearest the borehole is negligible because the electrical conductivity method samples rock extending to more than a metre from the borehole. A large number of laboratory measurements have been analysed in order to ensure that other effects that cannot be controlled in the in situ measurements do not influence the down-hole data. No unexpected effects have been found. Rock matrix porosity in situ measurements are extremely scarce. However, it has been possible to use some of the in situ measurements to estimate the increase in porosity when taking up rock from its natural environment to the laboratory. One example of such an investigation is briefly discussed to show how this was done. In one in situ diffusion experiment performed at a depth of 60 m in granitic rock in Sweden the experimental conditions were such that it was ensured that any rock stress changes due to the presence of the drift and the presence of the borehole were avoided. The rock was thus subject to 'virgin stress'. Over-coring after exposure to tracers for three and a half years and detailed sampling and

SR-Can. Data and uncertainty assessment. Matrix diffusivity and porosity in situ

Energy Technology Data Exchange (ETDEWEB)

Jinsong Liu; Loefgren, Martin; Neretnieks, Ivars [Dept. of Chemical Engineering and Technology, Royal Inst. of Technology, Stockholm (Sweden)

2006-12-15

and show beyond doubt that the electrical conductivity method using AC gives the expected information on transport properties (diffusivities) of the pores of crystalline rocks. The electrical conductivity method is much faster and can be used to measure large samples. It has recently been adopted for use in deep boreholes. Tens of thousands of measurements have been made at Simpevarp, Laxemar and Forsmark (Swedish sites) at depths of more than 1,000 m. The results of these measurements form the basis for our proposed diffusion values to be used in Performance Assessment (PA) for the candidate sites (Performance Assessment (PA) used in this report is synonymous to Safety Assessment (SA) sometimes used by other authors). The in situ data are obtained essentially in undisturbed rock and have not been subject to either stress release or disturbances due to sample preparation. The small disturbance nearest the borehole is negligible because the electrical conductivity method samples rock extending to more than a metre from the borehole. A large number of laboratory measurements have been analysed in order to ensure that other effects that cannot be controlled in the in situ measurements do not influence the down-hole data. No unexpected effects have been found. Rock matrix porosity in situ measurements are extremely scarce. However, it has been possible to use some of the in situ measurements to estimate the increase in porosity when taking up rock from its natural environment to the laboratory. One example of such an investigation is briefly discussed to show how this was done. In one in situ diffusion experiment performed at a depth of 60 m in granitic rock in Sweden the experimental conditions were such that it was ensured that any rock stress changes due to the presence of the drift and the presence of the borehole were avoided. The rock was thus subject to 'virgin stress'. Over-coring after exposure to tracers for three and a half years and detailed sampling

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

Regional and interspecific variation in Sr, Ca, and Sr/Ca ratios in avian eggshells from the USA.

Science.gov (United States)

Mora, Miguel A; Brattin, Bryan; Baxter, Catherine; Rivers, James W

2011-08-01

To examine regional variation in strontium (Sr), which at high concentrations may reduce eggshell quality, increase egg breakage and reproductive failure, we analyzed Sr, and calcium (Ca) concentrations and Sr/Ca ratios in eggshells from 20 avian species from California, Texas, Idaho, Kansas, and Michigan. In addition, we included data previously reported from Arizona to expand the regional comparisons and to better establish patterns of Sr, and Sr/Ca ratios in bird species across the United States. We found Sr concentrations varied significantly among regions, among species, and among foraging guilds; this variability is strongly influenced by the Sr/Ca ratios in surface water from locations close to the region where the eggshells were collected. Sr concentrations and Sr/Ca ratios were significantly higher in bird eggshells from the Volta wildlife region in the San Joaquin Valley, California and in various locales from Arizona. Sr concentrations and Sr/Ca ratios in bird eggshells from other locations in the USA were lower than those detected in these two regions. Among foraging guilds, invertivores had the highest Sr concentrations and Sr/Ca ratios and carnivores had the lowest. In general, the Sr/Ca ratio increased strongly with increasing Sr concentrations (R(2) = 0.99, P eggshells suggesting that these values could be determined from Sr/Ca ratios in water. Eggshell thickness was poorly correlated with Sr (R(2) = 0.03) but had a significant and positive correlation with Ca and was more properly correlated by a quadratic equation (R(2) = 0.50, Thickness = 2.13 - 0.02Ca - 3.07 * 10(-5)Ca(2)). Our study provides further evidence that Sr accumulates significantly in the avian eggshell, in some regions at concentrations which could be of concern for potential negative effects on reproduction. We suggest that when assessing the effects of metals on avian reproduction in regions with high Sr deposits in rock and soil, Sr concentrations in the eggshell also should be

Assessment of occupational exposure due to external sources of radiation. Safety guide

International Nuclear Information System (INIS)

2000-01-01

Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

Assessment of occupational exposure due to external sources of radiation. Safety guide

International Nuclear Information System (INIS)

2004-01-01

Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

Assessment of occupational exposure due to external sources of radiation. Safety guide

International Nuclear Information System (INIS)

1999-01-01

Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

Assessment of Sr-90 in water samples: precision and accuracy

Energy Technology Data Exchange (ETDEWEB)

Nisti, Marcelo B.; Saueia, Cátia H.R.; Castilho, Bruna; Mazzilli, Barbara P., E-mail: mbnisti@ipen.br, E-mail: chsaueia@ipen.br, E-mail: bcastilho@ipen.br, E-mail: mazzilli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The study of artificial radionuclides dispersion into the environment is very important to control the nuclear waste discharges, nuclear accidents and nuclear weapons testing. The accidents in Fukushima Daiichi Nuclear Power Plant and Chernobyl Nuclear Power Plant, released several radionuclides in the environment by aerial deposition and liquid discharge, with various level of radioactivity. The {sup 90}Sr was one of the elements released into the environment. The {sup 90}Sr is produced by nuclear fission with a physical half-life of 28.79 years with decay energy of 0.546 MeV. The aims of this study are to evaluate the precision and accuracy of three methodologies for the determination of {sup 90}Sr in water samples: Cerenkov, LSC direct method and with radiochemical separation. The performance of the methodologies was evaluated by using two scintillation counters (Quantulus and Hidex). The parameters Minimum Detectable Activity (MDA) and Figure Of Merit (FOM) were determined for each method, the precision and accuracy were checked using {sup 90}Sr standard solutions. (author)

A generic standard for assessing and managing activities with significant risk to health and safety

International Nuclear Information System (INIS)

Wilde, T.S.; Sandquist, G.M.

2005-01-01

Some operations and activities in industry, business, and government can present an unacceptable risk to health and safety if not performed according to established safety practices and documented procedures. The nuclear industry has extensive experience and commitment to assessing and controlling such risks. This paper provides a generic standard based upon DOE Standard DOE-STD-3007- 93, Nov 1993, Change Notice No. 1, Sep 1998. This generic standard can be used to assess practices and procedures employed by any industrial and government entity to ensure that an acceptable level of safety and control prevail for such operations. When any activity and operation is determined to involve significant risk to health and safety to workers or the public, the organization should adopt and establish an appropriate standard and methodology to ensure that adequate health and safety prevail. This paper uses DOE experience and standards to address activities with recognized potential for impact upon health and safety. Existing and future assessments of health and safety issues can be compared and evaluated against this generic standard for insuring that proper planning, analysis, review, and approval have been made. (authors)

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)