Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

Selection of important initiating events for Level 1 probabilistic safety assessment study at Puspati TRIGA Reactor

International Nuclear Information System (INIS)

Maskin, M.; Charlie, F.; Hassan, A.; Prak Tom, P.; Ramli, Z.; Mohamed, F.

2016-01-01

Highlights: • Identifying possible important initiating events (IEs) for Level 1 probabilistic safety assessment performed on research nuclear reactor. • Methods in screening and grouping IEs are addressed. • Focusing only on internal IEs due to random failures of components. - Abstract: This paper attempts to present the results in identifying possible important initiating events (IEs) as comprehensive as possible to be applied in the development of Level-1 probabilistic safety assessment (PSA) study. This involves the approaches in listing and the methods in screening and grouping IEs, by focusing only on the internal IEs due to random failures of components and human errors with full power operational conditions and reactor core as the radioactivity source. Five approaches were applied in listing the IEs and each step of the methodology was described and commented. The criteria in screening and grouping the IEs were also presented. The results provided the information on how the Malaysian PSA team applied the approaches in selecting the most probable IEs as complete as possible in order to ensure the set of IEs was identified systematically and as representative as possible, hence providing confidence to the completeness of the PSA study. This study is perhaps one of the first to address classic comprehensive steps in identifying important IEs to be used in a Level-1 PSA study.

Patient safety initiatives in Central and Eastern Europe: A mixed methods approach by the LINNEAUS collaboration on patient safety in primary care

Science.gov (United States)

Godycki-Cwirko, Maciek; Esmail, Aneez; Dovey, Susan; Wensing, Michel; Parker, Dianne; Kowalczyk, Anna; Błaszczyk, Honorata; Kosiek, Katarzyna

2015-01-01

ABSTRACT Background: Despite patient safety being recognized as an important healthcare issue in the European Union, there has been variable implementation of patient safety initiatives in Central and Eastern Europe (CEE). Objective: To assess the status of patient safety initiatives in countries in CEE; to describe a process of engagement in Poland, which can serve as a template for the implementation of patient safety initiatives in primary care. Methods: A mixed methods design was used. We conducted a review of literature focusing on publications from CEE, an inventory of patient safety initiatives in CEE countries, interviews with key informants, international survey, review of national reporting systems, and pilot demonstrator project in Poland with implementation of patient safety toolkits assessment. Results: There was no published patient safety research from Albania, Belarus, Greece, Latvia, Lithuania, Romania, or Russia. Nine papers were found from Bulgaria, Croatia, the Czech Republic, Poland, Serbia, and Slovenia. In most of the CEE countries, patient safety had been addressed at the policy level although the focus was mainly in hospital care. There was a dearth of activity in primary care. The use of patient improvement strategies was low. Conclusion: International cooperation as exemplified in the demonstrator project can help in the development and implementation of patient safety initiatives in primary care in changing the emphasis away from a blame culture to one where greater emphasis is placed on improvement and learning. PMID:26339839

Environment, safety and health progress assessment manual

International Nuclear Information System (INIS)

1992-12-01

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

Implementing Patient Safety Initiatives in Rural Hospitals

Science.gov (United States)

Klingner, Jill; Moscovice, Ira; Tupper, Judith; Coburn, Andrew; Wakefield, Mary

2009-01-01

Implementation of patient safety initiatives can be costly in time and energy. Because of small volumes and limited resources, rural hospitals often are not included in nationally driven patient safety initiatives. This article describes the Tennessee Rural Hospital Patient Safety Demonstration project, whose goal was to strengthen capacity for…

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

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

Children's safety initiative: a national assessment of pediatric educational needs among emergency medical services providers.

Science.gov (United States)

Hansen, Matthew; Meckler, Garth; Dickinson, Caitlyn; Dickenson, Kathryn; Jui, Jonathan; Lambert, William; Guise, Jeanne-Marie

2015-01-01

Emergency medical services (EMS) providers may have critical knowledge gaps in pediatric care due to lack of exposure and training. There is currently little evidence to guide educators to the knowledge gaps that most need to be addressed to improve patient safety. The objective of this study was to identify educational needs of EMS providers related to pediatric care in various domains in order to inform development of curricula. The Children's Safety Initiative-EMS performed a three-phase Delphi survey on patient safety in pediatric emergencies among providers and content experts in pediatric emergency care, including physicians, nurses, and prehospital providers of all levels. Each round included questions related to educational needs of providers or the effect of training on patient safety events. We identified knowledge gaps in the following domains: case exposure, competency and knowledge, assessment and decision making, and critical thinking and proficiency. Individual knowledge gaps were ranked by portion of respondents who ranked them "highly likely" (Likert-type score 7-10 out of 10) to contribute to safety events. There were 737 respondents who were included in analysis of the first phase of the survey. Paramedics were 50.8% of respondents, EMT-basics/first responders were 22%, and physicians 11.4%. The top educational priorities identified in the final round of the survey include pediatric airway management, responder anxiety when working with children, and general pediatric skills among providers. The top three needs in decision-making include knowing when to alter plans mid-course, knowing when to perform an advanced airway, and assessing pain in children. The top 3 technical or procedural skills needs were pediatric advanced airway, neonatal resuscitation, and intravenous/intraosseous access. For neonates, specific educational needs identified included knowing appropriate vital signs and preventing hypothermia. This is the first large-scale Delphi

Methodology for Safety Assessment Applied to Predisposal Waste Management. Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) 2004–2010)

International Nuclear Information System (INIS)

2015-12-01

Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) (2004–2010) The IAEA’s progamme on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) focused on approaches and mechanisms for application of safety assessment methodologies for the predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts, which have since been incorporated into IAEA Safety Standards Series No. GSG-3, Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste. In 2005, an initial specification was developed for the Safety Assessment Framework (SAFRAN) software tool to apply the SADRWMS flowcharts. In 2008, an in-depth application of the SAFRAN tool and the SADRWMS methodology was carried out on the predisposal management facilities of the Thailand Institute of Nuclear Technology Radioactive Waste Management Centre (TINT Facility). This publication summarizes the content and outcomes of the SADRWMS programme. The Chairman’s Report of the SADRWMS Project and the Report of the TINT test case are provided on the CD-ROM which accompanies this report

Integrated safety assessment report: Integrated Safety Assessment Program: Millstone Nuclear Power Station, Unit 1 (Docket No. 50-245): Draft report

International Nuclear Information System (INIS)

1987-04-01

The Integrated Safety Assessment Program (ISAP) was initiated in November 1984, by the US Nuclear Regulatory Commission to conduct integrated assessments for operating nuclear power reactors. The integrated assessment is conducted in a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. In addition, procedures will be established to allow for a periodic updating of the schedules to account for licensing issues that arise in the future. This report documents the review of Millstone Nuclear Power Station, Unit No. 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit No. 1, is one of two plants being reviewed under the pilot program for ISAP. This report indicates how 85 topics selected for review were addressed. This report presents the staff's recommendations regarding the corrective actions to resolve the 85 topics and other actions to enhance plant safety. The report is being issued in draft form to obtain comments from the licensee, nuclear safety experts, and the Advisory Committee for Reactor Safeguards (ACRS). Once those comments have been resolved, the staff will present its positions, along with a long-term implementation schedule from the licensee, in the final version of this report

Probabilistic safety assessment goals in Canada

International Nuclear Information System (INIS)

Snell, V.G.

1986-01-01

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

An evaluation of safety culture initiatives at BNSF Railway

Science.gov (United States)

2015-04-01

Major safety culture (SC) initiatives initiated in the FRA Office of Research, Technology and Development (RT&D), such as Clear Signal for Action (CSA), the Investigation of Safety Related Occurrences Protocol (ISROP), the Participative Safety Rules ...

Radiation safety without borders initiative

International Nuclear Information System (INIS)

Dibblee, Martha; Dickson, Howard; Krieger, Ken; Lopez, Jose; Waite, David; Weaver, Ken

2008-01-01

The Radiation Safety Without Borders (RSWB) initiative provides peer support to radiation safety professionals in developing countries, which bolsters the country's infrastructure and may lead the way for IRPA Associate membership. The Health Physics Society (HPS) recognizes that many nations do not possess the infrastructure to adequately control and beneficially use ionizing radiation. In a substantial number of countries, organized radiation protection programs are minimal. The RSWB initiative relies on HPS volunteers to assist their counterparts in developing countries with emerging health physics and radiation safety programs, but whose resources are limited, to provide tools that promote and support infrastructure and help these professionals help themselves. RSWB experience to date has shown that by providing refurbished instruments, promoting visits to a HPS venue, or visiting a country just to look provide valuable technical and social infrastructure experiences often missing in the developing nation's cadre of radiation safety professionals. HPS/RSWB with the assistance of the International Atomic Energy Agency (IAEA) pairs chapters with a country, with the expectation that the country's professional radiation safety personnel will form a foreign HPS chapter, and the country eventually will become an IRPA Associate. Although still in its formative stage, RSWB nonetheless has gotten valuable information in spite of the small number of missions. The RSWB initiative continues to have significant beneficial impacts, including: Improving the radiation safety infrastructure of the countries that participate; Assisting those countries without professional radiation safety societies to form one; Strengthening the humanitarian efforts of the United States; Enhancing Homeland Security efforts through improved control of radioactive material internationally. Developing countries, including those in Latin America, underwritten by IAEA, may take advantage of resources

MAPLE-X10 reactor safety assessment

International Nuclear Information System (INIS)

Cotnam, K.D.; Lounsbury, R.I.; Gillespie, G.E.

1990-01-01

This paper reports on the safety assessment of the 10 MW MAPLE-X10 reactor which has involved a substantial component of PSA analysis to supplement deterministic analysis. Initiating events are identified through the use of a master logic diagram. The events are then examined through event sequence diagrams, at the concept design stage, followed by a set of reliability analyses that are coordinated with the event sequence diagrams. Improvements identified through the reliability analyses are incorporated into the design to ensure that safety objectives are attained

Complementary assessment of the safety of French nuclear power plants

International Nuclear Information System (INIS)

Camarcat, N.; Pouget-Abadie, X.

2011-01-01

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

L-Band Digital Aeronautical Communications System Engineering - Initial Safety and Security Risk Assessment and Mitigation

Science.gov (United States)

Zelkin, Natalie; Henriksen, Stephen

2011-01-01

This document is being provided as part of ITT's NASA Glenn Research Center Aerospace Communication Systems Technical Support (ACSTS) contract NNC05CA85C, Task 7: "New ATM Requirements--Future Communications, C-Band and L-Band Communications Standard Development." ITT has completed a safety hazard analysis providing a preliminary safety assessment for the proposed L-band (960 to 1164 MHz) terrestrial en route communications system. The assessment was performed following the guidelines outlined in the Federal Aviation Administration Safety Risk Management Guidance for System Acquisitions document. The safety analysis did not identify any hazards with an unacceptable risk, though a number of hazards with a medium risk were documented. This effort represents a preliminary safety hazard analysis and notes the triggers for risk reassessment. A detailed safety hazards analysis is recommended as a follow-on activity to assess particular components of the L-band communication system after the technology is chosen and system rollout timing is determined. The security risk analysis resulted in identifying main security threats to the proposed system as well as noting additional threats recommended for a future security analysis conducted at a later stage in the system development process. The document discusses various security controls, including those suggested in the COCR Version 2.0.

Electronuclear's safety culture assessment and enhancement program

International Nuclear Information System (INIS)

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

2002-01-01

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

Environment, safety and health progress assessment manual

International Nuclear Information System (INIS)

1992-12-01

On June 27, 1989, the Secretary of Energy announced a 10-Point Initiative to strengthen environment, safety, and health (ES ampersand H) programs, and waste management activities at DOE production, research, and testing facilities. One of the points involved conducting dent Tiger Team Assessments of DOE operating facilities. The Office of Special independent Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This manual documents the processes to be used to perform the ES ampersand H Progress Assessments. It was developed based upon the lessons learned from Tiger Team Assessments, the two pilot Progress Assessments, and Progress Assessments that have been completed. The manual will be updated periodically to reflect lessons learned or changes in policy

Safety analysis and risk assessment handbook

International Nuclear Information System (INIS)

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

1997-01-01

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

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

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

The safety net medical home initiative: transforming care for vulnerable populations.

Science.gov (United States)

Sugarman, Jonathan R; Phillips, Kathryn E; Wagner, Edward H; Coleman, Katie; Abrams, Melinda K

2014-11-01

Despite findings that medical homes may reduce or eliminate health care disparities among underserved and minority populations, most previous medical home pilot and demonstration projects have focused on health care delivery systems serving commercially insured patients and Medicare beneficiaries. To develop a replicable approach to support medical home transformation among diverse practices serving vulnerable and underserved populations. Facilitated by a national program team, convening organizations in 5 states provided coaching and learning community support to safety net practices over a 4-year period. To guide transformation, we developed a framework of change concepts aligned with supporting tools including implementation guides, activity checklists, and measurement instruments. Sixty-five health centers, homeless clinics, private practices, residency training centers, and other safety net practices in Colorado, Idaho, Massachusetts, Oregon, and Pennsylvania. We evaluated implementation of the change concepts using the Patient-Centered Medical Home-Assessment, and conducted a survey of participating practices to assess perceptions of the impact of the technical assistance. All practices implemented key features of the medical home model, and nearly half (47.6%) implemented the 33 identified key changes to a substantial degree as evidenced by level A Patient-Centered Medical Home-Assessment scores. Two thirds of practices that achieved substantial implementation did so only after participating in the initiative for >2 years. By the end of the initiative, 83.1% of sites achieved external recognition as medical homes. Despite resource constraints and high-need populations, safety net clinics made considerable progress toward medical home implementation when provided robust, multimodal support over a 4-year period.

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

International Nuclear Information System (INIS)

Kot, C. A.

1998-01-01

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

SPI Project Criticality Task Force initial review and assessment

International Nuclear Information System (INIS)

McKinley, K.B.; Cannon, J.W.; Marsden, R.S.; Worle, H.A.

1980-03-01

The Slagging Pyrolysis Incinerator (SPI) Facility is being developed to process transuranic waste stored and buried at the Idaho National Engineering Laboratory (INEL) into a chemically inert, physically stable, basalt-like residue acceptable for a Federal Repository. A task force was established by the SPI Project Division to review and assess all aspects of criticality safety for the SPI Facility. This document presents the initial review, evaluations, and recommendations of the task force and includes the following: background information on waste characterization, and criticality control approaches and philosophies, a description of the SPI Facility Waste Processing Building, a review and assessment of potentially relevant codes and regulations; a review and assessment of the present state of criticality and assaying/monitoring studies, and recommendations for changes in and additions to these studies. The review and assessment of potentially relevant codes and regulations indicate that ERDAM 0530, Nuclear Criticality Safety should be the controlling document for criticality safety for the SPI Project. In general, the criticality control approaches and philosophies for the SPI Project comply with this document

Integrated safety assessment report, Haddam Neck Plant (Docket No. 50-213): Integrated Safety Assessment Program: Draft report

International Nuclear Information System (INIS)

1987-07-01

The integrated assessment is conducted on a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. Procedures allow for a periodic updating of the schedules to account for licensing issues that arise in the future. The Haddam Neck Plant is one of two plants being reviewed under the pilot program. This report indicates how 82 topics selected for review were addressed, and presents the staff's recommendations regarding the corrective actions to resolve the 82 topics and other actions to enhance plant safety. 135 refs., 4 figs., 5 tabs

Determination of Initial Conditions for the Safety Analysis by Random Sampling of Operating Parameters

International Nuclear Information System (INIS)

Jeong, Hae-Yong; Park, Moon-Ghu

2015-01-01

In most existing evaluation methodologies, which follow a conservative approach, the most conservative initial conditions are searched for each transient scenario through tremendous assessment for wide operating windows or limiting conditions for operation (LCO) allowed by the operating guidelines. In this procedure, a user effect could be involved and a remarkable time and human resources are consumed. In the present study, we investigated a more effective statistical method for the selection of the most conservative initial condition by the use of random sampling of operating parameters affecting the initial conditions. A method for the determination of initial conditions based on random sampling of plant design parameters is proposed. This method is expected to be applied for the selection of the most conservative initial plant conditions in the safety analysis using a conservative evaluation methodology. In the method, it is suggested that the initial conditions of reactor coolant flow rate, pressurizer level, pressurizer pressure, and SG level are adjusted by controlling the pump rated flow, setpoints of PLCS, PPCS, and FWCS, respectively. The proposed technique is expected to contribute to eliminate the human factors introduced in the conventional safety analysis procedure and also to reduce the human resources invested in the safety evaluation of nuclear power plants

Results of the CANDU 3 probabilistic safety assessment

Energy Technology Data Exchange (ETDEWEB)

Jaitly, R K [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10{sup -6}/year. 4 refs., 1 fig.

Results of the CANDU 3 probabilistic safety assessment

International Nuclear Information System (INIS)

Jaitly, R.K.

1995-01-01

The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10 -6 /year. 4 refs., 1 fig

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)

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.

Initial assessment of MHTGR confinement releases

International Nuclear Information System (INIS)

Maneke, J.L.; Lanning, D.D.; Lidsky, L.M.

1986-01-01

Initial investigation of Modular High Temperature Gas-Cooled Reactor (MHTGR) designs suggest that source terms during postulated accidents will be considerably lower than Light Water Reactor (LWR) source term estimates. These lower postulated accident releases are not only a safety incentive, but also an economic incentive for the development of this reactor type. For example, it is hoped that a filtered confinement building, rather than a more expensive LWR-like containment building, would adequately protect the public from radiological releases. The ability of a confinement building to satisfy safety requirements for the MHTGR depends on several reactor parameters, such as fuel quality, reactor design, and the design of the reactor building. SCIMCA, a Simple Code for Initial MHTGR Confinement Assessment has been developed for preliminary MHTGR building requirement calculations. The code is capable of modeling a decay chain with a maximum of five regions. Phenomena such as fission product decay and buildup, natural deposition, building filtration, and intercompartmental transport are incorporated. SCIMCA models reduction mechanisms, such as dispersion and decay, occurring as radioactivity is transported through the environment. A subroutine for calculating doses at specified distances has also been included

A study on the methodology of probabilistic safety assessment for KALIMER

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwan Seong; Kwon, Young Min; Lee, Yong Bum; Jeong, Hae Yong; Yang, Joon Eon; Ha, Kyu Suk; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

Existing Probabilistic Safety Assessment(PSA) is a method for Light Water Reactor or Pressurized Heavy Water Reactor. Because KALIMER is different from these reactor, the new methodology of PSA need to be developed. In this paper, the PSA of Power Reactor Inherently Safety Module(PRISM) is analyzed, and Initiating Event such as Experiential Assessment, Logical Assessment and Failure Mode Effect Analysis(FMEA) is reviewed. Also, Pipe Damage Frequency Method is suggested for KALIMER. And the Reliability Physical method of Passive System, which is a chief safety system of KALIMER, is reviewed and its applicability is investigated. Finally, for the Preliminary PSA of KALIMER, Intermediate Heat Transfer System is analyzed. 23 refs., 10 figs., 13 tabs. (Author)

76 FR 29773 - Call for Participation in Pillbox Patient-Safety Initiative

Science.gov (United States)

2011-05-23

... Pillbox Patient-Safety Initiative ACTION: Notice. SUMMARY: The National Library of Medicine (NLM) invites..., production version of Pillbox. This initiative is an important element of ongoing efforts to enhance patient.... SUPPLEMENTARY INFORMATION: NLM has established Pillbox, an initiative to enhance patient safety, by making...

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

DOE spent nuclear fuel -- Nuclear criticality safety challenges and safeguards initiatives

International Nuclear Information System (INIS)

Hopper, C.M.

1994-01-01

The field of nuclear criticality safety is confronted with growing technical challenges and the need for forward-thinking initiatives to address and resolve issues surrounding economic, safe and secure packaging, transport, interim storage, and long-term disposal of spent nuclear fuel. These challenges are reflected in multiparameter problems involving optimization of packaging designs for maximizing the density of material per package while ensuring subcriticality and safety under variable normal and hypothetical transport and storage conditions and for minimizing costs. Historic and recently revealed uncertainties in basic data used for performing nuclear subcriticality evaluations and safety analyses highlight the need to be vigilant in assessing the validity and range of applicability of calculational evaluations that represent extrapolations from ''benchmark'' data. Examples of these uncertainties are provided. Additionally, uncertainties resulting from the safeguarding of various forms of fissionable materials in transit and storage are discussed

Safety assessment for the IS process in a hydrogen production facility

International Nuclear Information System (INIS)

Cho, Nam Chul

2005-08-01

A substitute energy development have been required due to the dry up of the fossil fuel and an environmental problem. Consequently, among substitute energy to be discussed, producing hydrogen from water which does not release carbon is a very promising technology. Also, Iodine-Sulfur(IS) thermochemical water decomposition is one of the promising process which is used to produce hydrogen efficiently using the high temperature gas-cooled reactor(HTGR) as an energy source that is possible to supply heat over 1000 .deg. C. In this study, to make a safety assessment of the hydrogen production using the IS process, an initiating events analysis and an accident scenario modeling considering the relief system were carried out. A method for initiating event identification used the Master Logic Diagram(MLD) that is logical and deductive. As a result, 9 initiating events that cause a leakage of the chemical material were identified. 6 accident scenario based on the initiating event are identified and quantified to the event trees. The frequency of the chemical material leakage produced by IS process is estimated relatively high to the value of 1.22x10 -4 /y. Therefore, it requires more effort on safety of the hydrogen production which can be considered as a part of the nuclear system and safety management research to increase social acceptability. Moreover, these methods will be helpful to the safety assessment of the hydrogen production system of the IS process in general

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 different fracture-mechanical J-integral initiation values with regard to their usability in the safety assessment of components

International Nuclear Information System (INIS)

Eisele, U.; Roos, E.

1991-01-01

Determining fracture-mechanical material characteristic values on the basis of the J-integral is described and stipulated in a variety of standards and guidelines. The individual specifications differ in terms of procedure when determining the characteristic values and, therefore, also in terms of the meaningfulness of the results. This paper presents the different procedures, suggested in the course of the development of test methods in the field of elastic-plastic fracture mechanics, used to characterize crack initiation behaviour with regard to their features as material characteristic values and their usability in the safety assessment of components. (orig.)

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

Surgical Safety Training of World Health Organization Initiatives.

Science.gov (United States)

Davis, Christopher R; Bates, Anthony S; Toll, Edward C; Cole, Matthew; Smith, Frank C T; Stark, Michael

2014-01-01

Undergraduate training in surgical safety is essential to maximize patient safety. This national review quantified undergraduate surgical safety training. Training of 2 international safety initiatives was quantified: (1) World Health Organization (WHO) "Guidelines for Safe Surgery" and (2) Department of Health (DoH) "Principles of the Productive Operating Theatre." Also, 13 additional safety skills were quantified. Data were analyzed using Mann-Whitney U tests. In all, 23 universities entered the study (71.9% response). Safety skills from WHO and DoH documents were formally taught in 4 UK medical schools (17.4%). Individual components of the documents were taught more frequently (47.6%). Half (50.9%) of the additional safety skills identified were taught. Surgical societies supplemented safety training, although the total amount of training provided was less than that in university curricula (P < .0001). Surgical safety training is inadequate in UK medical schools. To protect patients and maximize safety, a national undergraduate safety curriculum is recommended. © 2013 by the American College of Medical Quality.

Assessment of Human Performance and Safety Culture at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Toth, Janos; Hadnagy, Lajos

2002-01-01

Evaluation of human performance and safety culture of the personnel at a Nuclear Power Plant is a very important element of the self assessment process. At the Paks NPP a systematic approach to this problem started in the early 90's. The first comprehensive analysis of the human performance of the personnel was performed by the Hungarian Research Institute for Electric Power (VEIKI). The analysis of human failures is also a part of the investigation and analysis of safety related reported events. This human performance analysis of events is carried out by the Laboratory of Psychology of the plant and a supporting organisation namely the Department of Ergonomics and Psychology of the Budapest University of Technical and Economical Sciences. The analysis of safety culture at the Paks NPP has been in the focus of attention since the implementation of the INSAG-4 document started world-wide. In 1993 an IAEA model project namely 'Strengthening Training for Operational Safety' was initiated with a sub-project called 'Enhancement of Safety Culture'. Within this project the first step was the initial assessment of the safety culture level at the Paks NPP. It was followed by some corrective actions and safety culture improvement programme. In 1999 the second assessment was performed in order to evaluate the progress as a result of the improvement programme. A few indicators reflecting the elements of safety culture were defined and compared. The assessment of the safety culture with a survey among the managers was performed in September 2000 and the results are being evaluated at the moment. The intention of the plant management is to repeat the assessment every 2-3 years and evaluate the trend of the indicator. (authors)

Safety assessment of HLW geological disposal system

International Nuclear Information System (INIS)

Naito, Morimasa

2006-01-01

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

Software Dependability and Safety Evaluations ESA's Initiative

Science.gov (United States)

Hernek, M.

ESA has allocated funds for an initiative to evaluate Dependability and Safety methods of Software. The objectives of this initiative are; · More extensive validation of Safety and Dependability techniques for Software · Provide valuable results to improve the quality of the Software thus promoting the application of Dependability and Safety methods and techniques. ESA space systems are being developed according to defined PA requirement specifications. These requirements may be implemented through various design concepts, e.g. redundancy, diversity etc. varying from project to project. Analysis methods (FMECA. FTA, HA, etc) are frequently used during requirements analysis and design activities to assure the correct implementation of system PA requirements. The criticality level of failures, functions and systems is determined and by doing that the critical sub-systems are identified, on which dependability and safety techniques are to be applied during development. Proper performance of the software development requires the development of a technical specification for the products at the beginning of the life cycle. Such technical specification comprises both functional and non-functional requirements. These non-functional requirements address characteristics of the product such as quality, dependability, safety and maintainability. Software in space systems is more and more used in critical functions. Also the trend towards more frequent use of COTS and reusable components pose new difficulties in terms of assuring reliable and safe systems. Because of this, its dependability and safety must be carefully analysed. ESA identified and documented techniques, methods and procedures to ensure that software dependability and safety requirements are specified and taken into account during the design and development of a software system and to verify/validate that the implemented software systems comply with these requirements [R1].

Safety and Security Interface Technology Initiative

Energy Technology Data Exchange (ETDEWEB)

Dr. Michael A. Lehto; Kevin J. Carroll; Dr. Robert Lowrie

2007-05-01

Safety and Security Interface Technology Initiative Mr. Kevin J. Carroll Dr. Robert Lowrie, Dr. Micheal Lehto BWXT Y12 NSC Oak Ridge, TN 37831 865-576-2289/865-241-2772 carrollkj@y12.doe.gov Work Objective. Earlier this year, the Energy Facility Contractors Group (EFCOG) was asked to assist in developing options related to acceleration deployment of new security-related technologies to assist meeting design base threat (DBT) needs while also addressing the requirements of 10 CFR 830. NNSA NA-70, one of the working group participants, designated this effort the Safety and Security Interface Technology Initiative (SSIT). Relationship to Workshop Theme. “Supporting Excellence in Operations Through Safety Analysis,” (workshop theme) includes security and safety personnel working together to ensure effective and efficient operations. One of the specific workshop elements listed in the call for papers is “Safeguards/Security Integration with Safety.” This paper speaks directly to this theme. Description of Work. The EFCOG Safety Analysis Working Group (SAWG) and the EFCOG Security Working Group formed a core team to develop an integrated process involving both safety basis and security needs allowing achievement of the DBT objectives while ensuring safety is appropriately considered. This effort garnered significant interest, starting with a two day breakout session of 30 experts at the 2006 Safety Basis Workshop. A core team was formed, and a series of meetings were held to develop that process, including safety and security professionals, both contractor and federal personnel. A pilot exercise held at Idaho National Laboratory (INL) in mid-July 2006 was conducted as a feasibility of concept review. Work Results. The SSIT efforts resulted in a topical report transmitted from EFCOG to DOE/NNSA in August 2006. Elements of the report included: Drivers and Endstate, Control Selections Alternative Analysis Process, Terminology Crosswalk, Safety Basis

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

Safety Assessment in the AREVA Group: Operating Experience from a Self-Assessment Tool

International Nuclear Information System (INIS)

Coye de Brunélis, T.; Mignot, E.; Sidaner, J.-F.

2016-01-01

The expression “safety culture” first appeared following analysis of the Chernobyl accident in 1986. It was first defined in INSAG-4 (International Nuclear Safety Advisory Group safety series) in 1991. Other events have occurred in nuclear facilities and during transportation since Chernobyl: Tokai Mura in 1999, Roissy Transport in 2002, Davis Besse in 2002, Thorp in 2005. These events show that the initial approach was too simplistic. Based on this observation, the definition of safety culture was supplemented by including concepts of cultural value (associated with the country and the company) and human and organizational factors, and was integrated in that form with the emergence and implementation of integrated management systems (IMS). Today, the concept of nuclear safety culture covers a wide set of factors such as safety, quality, corporate culture, defined processes and policies, organizations and related resources. Any assessment of people’s safety culture, particularly people directly involved in facility operations, is thus part of a comprehensive policy and contributes to a de facto demonstration of the priority which management assigns to safety.

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-15

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

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)

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

A safety assessment of the SEAFP fuel cycle systems

International Nuclear Information System (INIS)

Natalizio, A.; Kalyanam, K.; Ciattaglia, S.; Pace, L. di

1995-01-01

CFFTP and ENEA participated in a joint safety assessment of the fuel cycle design developed for the SEAFP fusion power reactor study (SEAFP: Safety and Environmental Assessment of Fusion Power). The assessment considered both conventional (deflagation/detonation) and radioactive hazards associated with the handling of significant quantities of hydrogen isotopes (H, D and T). Accordingly, the assessment focused on systems or equipment where either the flow rate, or inventory, of hydrogen isotopes was large. A systematic and thorough assessment of initiating events that can lead to an accidental release of tritium into the environment was the first step of the analysis process. This review demonstrated that, in all cases, there are at least two lines of defence available for mitigating the consequences of such accidents -i.e., secondary confinement (glove box, second pipe, caisson, etc.) and the building confinement, backed-up by an air detritiation capability. Therefore, large releases of tritium to the environment will occur only at very low frequencies. (orig.)

Probabilistic safety assessment based expert systems in support of dynamic risk assessment

International Nuclear Information System (INIS)

Varde, P.V.; Sharma, U.L.; Marik, S.K.; Raina, V.K.; Tikku, A.C.

2006-01-01

Probabilistic Safety Assessment (PSA) studies are being performed, world over as part of integrated risk assessment for Nuclear Power Plants and in many cases PSA insight is utilized in support of decision making. Though the modern plants are built with inherent safety provisions, particularly to reduce the supervisory requirements during initial period into the accident, it is always desired to develop an efficient user friendly real-time operator advisory system for handling of plant transients/emergencies which would be of immense benefit for the enhancement of operational safety of the plant. This paper discusses an integrated approach for the development of operator support system. In this approach, PSA methodology and the insight obtained from PSA has been utilized for development of knowledge based or rule based experts system. While Artificial Neural Network (ANN) approach has been employed for transient identification, rule-base expert system shell environment was used for the development of diagnostic module in this system. Attempt has been made to demonstrate that this approach offers an efficient framework for addressing requirements related to handling of real-time/dynamic scenario. (author)

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

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

The Environmental Health/Home Safety Education Project: a successful and practical U.S.-Mexico border initiative.

Science.gov (United States)

Forster-Cox, Susan C; Mangadu, Thenral; Jacquez, Benjamín; Fullerton, Lynne

2010-05-01

The Environmental Health/Home Safety Education Project (Proyecto de Salud Ambiental y Seguridad en el Hogar) has been developed in response to a wide array of severe and often preventable environmental health issues occurring in and around homes on the U.S.-Mexico border. Utilizing well-trained community members, called promotoras , homes are visited and assessed for potential environmental hazards, including home fire and food safety issues. Data analyzed from project years 2002 to 2005 shows a significant impact in knowledge levels and initial behavior change among targeted participants as it relates to fire and food safety issues. Since the initiation of the project in 1999, hundreds of participants have improved their quality of life by making their homes safer. The project has proven to be sustainable, replicable, flexible, and attractive to funders.

Overview of waste isoltaion safety assessment program and description of source term characterization task at PNL

International Nuclear Information System (INIS)

Bradley, D.

1977-01-01

A project is being conducted to develop and illustrate the methods and obtain the data necessary to assess the safety of long-term disposal of high-level radioactive waste in geologic formations. The methods and data will initially focus on generic geologic isolation systems but will ultimately be applied to the long-term safety assessment of specific candidate sites that are selected in the NWTS Program. The activities of waste isolation safety assessment (WISAP) are divided into six tasks: (1) Safety Assessment Concepts and Methods, (2) Disruptive Event Analysis, (3) Source Characterization, (4) Transport Modeling, (5) Transport Data and (6) Societal Acceptance

OSART Independent Safety Culture Assessment (ISCA) Guidelines

International Nuclear Information System (INIS)

2016-01-01

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

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.)

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

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.

Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

Energy Technology Data Exchange (ETDEWEB)

Kinker, M., E-mail: M.Kinker@iaea.org [International Atomic Energy Agency (IAEA), Vienna (Austria); Avila, R.; Hofman, D., E-mail: rodolfo@facilia.se [FACILIA AB, Stockholm (Sweden); Jova Sed, L., E-mail: jovaluis@gmail.com [Centro Nacional de Seguridad Nuclear (CNSN), La Habana (Cuba); Ledroit, F., E-mail: frederic.ledroit@irsn.fr [IRSN PSN-EXP/SSRD/BTE, (France)

2013-07-01

In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

International Nuclear Information System (INIS)

Kinker, M.; Avila, R.; Hofman, D.; Jova Sed, L.; Ledroit, F.

2013-01-01

In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

Safety assessment principles for nuclear plants

International Nuclear Information System (INIS)

1992-01-01

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

Development of safety related technology and infrastructure for safety assessment

International Nuclear Information System (INIS)

Venkat Raj, V.

1997-01-01

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

Thinking of the safety assessment of HLW disposal

International Nuclear Information System (INIS)

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

2014-01-01

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

Terrain Safety Assessment in Support of the Mars Science Laboratory Mission

Science.gov (United States)

Kipp, Devin

2012-01-01

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

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)

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

Planning report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-06-01

This document is a planning report for SKB's next assessment of long-term safety for a KBS 3 repository. The assessment, SR-Can, is to be finished by the end of 2005 and will be used for SKB's application to build an Encapsulation plant for spent nuclear fuel. Apart from outlining the methodology, the report discusses the handling in SR-Can of a number of important issues regarding the near field, the geosphere, the biosphere, the climatic evolution etc. The Swedish nuclear safety and radiation protection authorities have recently issued regulations concerning the final disposal of nuclear waste. The principal compliance criterion states that the annual risk of harmful effects must not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk. There are also a number of requirements on methodological aspects of the safety assessment as well as on the contents of a safety report. The regulations are reproduced in an Appendix to this report. The primary safety function of the KBS 3 system is to completely isolate the spent nuclear fuel within copper canisters over the entire assessment period, which will be one million years in SR-Can. Should a canister be damaged, the secondary safety function is to retard any releases from the canisters. The main steps of the assessment are the following: 1. Qualitative system description, FEP processing: This step consists of defining a system boundary and of describing the system on a format suitable for the safety assessment. Databases of relevant features, events and processes influencing long-term safety are structured and used as one starting point for the assessment. 2. Initial state descriptions. 3. Process descriptions: In this step all identified processes within the system boundary involved in the long-term evolution of the system are described in detail. 4. Description of boundary conditions: This step is a broad description of the evolution of the boundaries of the system

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

International Nuclear Information System (INIS)

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

1983-08-01

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

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.

Safety Assessment Methodologies and Their Application in Development of Near Surface Waste Disposal Facilities--ASAM Project

International Nuclear Information System (INIS)

Batandjieva, B.; Metcalf, P.

2003-01-01

Safety of near surface disposal facilities is a primary focus and objective of stakeholders involved in radioactive waste management of low and intermediate level waste and safety assessment is an important tool contributing to the evaluation and demonstration of the overall safety of these facilities. It plays significant role in different stages of development of these facilities (site characterization, design, operation, closure) and especially for those facilities for which safety assessment has not been performed or safety has not been demonstrated yet and the future has not been decided. Safety assessments also create the basis for the safety arguments presented to nuclear regulators, public and other interested parties in respect of the safety of existing facilities, the measures to upgrade existing facilities and development of new facilities. The International Atomic Energy Agency (IAEA) has initiated a number of research coordinated projects in the field of development and improvement of approaches to safety assessment and methodologies for safety assessment of near surface disposal facilities, such as NSARS (Near Surface Radioactive Waste Disposal Safety Assessment Reliability Study) and ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) projects. These projects were very successful and showed that there is a need to promote the consistent application of the safety assessment methodologies and to explore approaches to regulatory review of safety assessments and safety cases in order to make safety related decisions. These objectives have been the basis of the IAEA follow up coordinated research project--ASAM (Application of Safety Assessment Methodologies for Near Surface Disposal Facilities), which will commence in November 2002 and continue for a period of three years

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

Geosphere process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Skagius, Kristina

2006-09-01

This report documents geosphere processes identified as relevant to the long-term safety of a KBS- repository, and forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The following excerpts describe the methodology, and clarify the role of this process report in the assessment. 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, will evolve over time. Future states of the system will depend on the initial state of the system, a number of radiation related, thermal, hydraulic, mechanical, chemical and biological processes acting within the repository system over time, and external influences acting on the system. A methodology in ten steps has been developed for SR-Can described below. Identification of factors to consider (FEP processing): This step consists of identifying all the factors that need to be included in the analysis. Experience from earlier safety assessments and KBS-specific and international databases of relevant features, events and processes influencing long-term safety are utilised. Based on the results of the FEP processing, an SR-Can FEP catalogue, containing FEPs to be handled in SR-Can, has been established. The initial state of the system is described based on the design specifications of the KBS repository, a descriptive model of the repository site and a site-specific layout of the repository. The initial state of the fuel and the engineered components is that immediately after deposition, as described in the SR-Can Initial state report. The initial state of the geosphere and the biosphere is that of the natural system prior to excavation, as described in the site descriptive models. The repository layouts adapted to the sites are provided in underground

Geosphere process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Skagius, Kristina [Kemakta Konsult AB, Stockholm (SE)] (ed.)

2006-09-15

This report documents geosphere processes identified as relevant to the long-term safety of a KBS- repository, and forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The following excerpts describe the methodology, and clarify the role of this process report in the assessment. 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, will evolve over time. Future states of the system will depend on the initial state of the system, a number of radiation related, thermal, hydraulic, mechanical, chemical and biological processes acting within the repository system over time, and external influences acting on the system. A methodology in ten steps has been developed for SR-Can described below. Identification of factors to consider (FEP processing): This step consists of identifying all the factors that need to be included in the analysis. Experience from earlier safety assessments and KBS-specific and international databases of relevant features, events and processes influencing long-term safety are utilised. Based on the results of the FEP processing, an SR-Can FEP catalogue, containing FEPs to be handled in SR-Can, has been established. The initial state of the system is described based on the design specifications of the KBS repository, a descriptive model of the repository site and a site-specific layout of the repository. The initial state of the fuel and the engineered components is that immediately after deposition, as described in the SR-Can Initial state report. The initial state of the geosphere and the biosphere is that of the natural system prior to excavation, as described in the site descriptive models. The repository layouts adapted to the sites are provided in underground

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)

Assessing and improving the safety culture of non-power nuclear installations

International Nuclear Information System (INIS)

Bastin, S.J.; Cameron, R.F.; McDonald, N.R.; Adams, A.; Williamson, A.

2000-01-01

The development and application of safety culture principles has understandably focused on nuclear power plant and fuel cycle facilities and has been based on studies in Europe, North America, Japan and Korea. However, most radiation injuries and deaths have resulted from the mishandling of radioactive sources, inadvertent over-exposure to X-rays and critically incidents, unrelated to nuclear power plant. Within the Forum on Nuclear Cooperation in Asia (FNCA), Australia has been promoting initiatives to apply safety culture principles across all nuclear and radiation application activities and in a manner that is culturally appropriate for Asian countries. ANSTO initiated a Safety Culture Project in 1996 to develop methods for assessing and improving safety culture at nuclear and radiation installations other than power reactors and to trial these at ANSTO and in the Asian region. The project has sensibly drawn on experience from the nuclear power industry, particularly in Japan and Korea. There has been a positive response in the participating countries to addressing safety culture issues in non-power nuclear facilities. This paper reports on the main achievements of the project. Further goals of the project are also identified. (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)

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

Science.gov (United States)

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

1993-01-01

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

Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

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

Safety assessments for potential exposures

International Nuclear Information System (INIS)

Dunn, D.I.

2012-04-01

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

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

International Nuclear Information System (INIS)

Song, Tae Young

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

Safety assessment for spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

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

A comparative review of patient safety initiatives for national health information technology

DEFF Research Database (Denmark)

Magrabi, Farah; Aarts, Jos; Nøhr, Christian

2013-01-01

OBJECTIVE: To collect and critically review patient safety initiatives for health information technology (HIT). METHOD: Publicly promulgated set of advisories, recommendations, guidelines, or standards potentially addressing safe system design, build, implementation or use were identified...... by searching the websites of regional and national agencies and programmes in a non-exhaustive set of exemplar countries including England, Denmark, the Netherlands, the USA, Canada and Australia. Initiatives were categorised by type and software systems covered. RESULTS: We found 27 patient safety initiatives...... were aimed at certification in the USA, Canada and Australia. Safety is addressed alongside interoperability in the Australian certification programme but it is not explicitly addressed in the US and Canadian programmes, though conformance with specific functionality, interoperability, security...

Leadership and Safety Management: Regulatory Initiatives for Enhancing Nuclear Safety in the Republic of Korea

International Nuclear Information System (INIS)

Yun, C.H.; Park, Y.W.; Choi, K.S.

2010-01-01

Since the construction of the first nuclear power plant (NPP) in the Republic of Korea in 1978, a high level of nuclear safety has continued to be maintained. This has been the important basis on which the continuous construction of NPPs has been possible in the country. To date, regulatory initiatives, leaderships and strategies adopting well harmonized regulatory systems and practices of advanced countries have contributed to improving the effectiveness and efficiency of safety regulation and further enhancing nuclear safety. The outcomes have resulted in a high level of safety and performance of Korean NPPs, attributing largely to the safety promotion policy. Recently, with the support of the Korean Ministry of Education, Science and Technology (MEST), the Korea Institute of Nuclear Safety (KINS) established the International Nuclear Safety School and created a Nuclear Safety Master's Degree Programme. Further, it developed multilateral and bilateral cooperation with other agencies to promote global nuclear safety, with the aim of providing knowledge and training to new entrant countries in establishing the safety infrastructure necessary for ensuring an acceptable level of nuclear safety. (author)

Outcomes of an international initiative for harmonization of low power and shutdown probabilistic safety assessment

Directory of Open Access Journals (Sweden)

Manna Giustino

2010-01-01

Full Text Available Many probabilistic safety assessment studies completed to the date have demonstrated that the risk dealing with low power and shutdown operation of nuclear power plants is often comparable with the risk of at-power operation, and the main contributors to the low power and shutdown risk often deal with human factors. Since the beginning of the nuclear power generation, human performance has been a very important factor in all phases of the plant lifecycle: design, commissioning, operation, maintenance, surveillance, modification, decommissioning and dismantling. The importance of this aspect has been confirmed by recent operating experience. This paper provides the insights and conclusions of a workshop organized in 2007 by the IAEA and the Joint Research Centre of the European Commission, on Harmonization of low power and shutdown probabilistic safety assessment for WWER nuclear power plants. The major objective of the workshop was to provide a comparison of the approaches and the results of human reliability analyses and gain insights in the enhanced handling of human factors.

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)

A Methodology for Safety Culture Impact Assessment

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

A Methodology for Safety Culture Impact Assessment

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2014-01-01

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

Safety and Security Interface Technology Initiative

International Nuclear Information System (INIS)

Dr. Michael A. Lehto; Kevin J. Carroll; Dr. Robert Lowrie

2007-01-01

Earlier this year, the Energy Facility Contractors Group (EFCOG) was asked to assist in developing options related to acceleration deployment of new security-related technologies to assist meeting design base threat (DBT) needs while also addressing the requirements of 10 CFR 830. NNSA NA-70, one of the working group participants, designated this effort the Safety and Security Interface Technology Initiative (SSIT). Relationship to Workshop Theme. ''Supporting Excellence in Operations Through Safety Analysis'', (workshop theme) includes security and safety personnel working together to ensure effective and efficient operations. One of the specific workshop elements listed in the call for papers is ''Safeguards/Security Integration with Safety''. This paper speaks directly to this theme. Description of Work. The EFCOG Safety Analysis Working Group (SAWG) and the EFCOG Security Working Group formed a core team to develop an integrated process involving both safety basis and security needs allowing achievement of the DBT objectives while ensuring safety is appropriately considered. This effort garnered significant interest, starting with a two day breakout session of 30 experts at the 2006 Safety Basis Workshop. A core team was formed, and a series of meetings were held to develop that process, including safety and security professionals, both contractor and federal personnel. A pilot exercise held at Idaho National Laboratory (INL) in mid-July 2006 was conducted as a feasibility of concept review. Work Results. The SSIT efforts resulted in a topical report transmitted from EFCOG to DOE/NNSA in August 2006. Elements of the report included: Drivers and Endstate, Control Selections Alternative Analysis Process, Terminology Crosswalk, Safety Basis/Security Documentation Integration, Configuration Control, and development of a shared ''tool box'' of information/successes. Specific Benefits. The expectation or end state resulting from the topical report and associated

Planning report for the safety assessment SR-Can

International Nuclear Information System (INIS)

2003-06-01

This document is a planning report for SKB's next assessment of long-term safety for a KBS 3 repository. The assessment, SR-Can, is to be finished by the end of 2005 and will be used for SKB's application to build an Encapsulation plant for spent nuclear fuel. Apart from outlining the methodology, the report discusses the handling in SR-Can of a number of important issues regarding the near field, the geosphere, the biosphere, the climatic evolution etc. The Swedish nuclear safety and radiation protection authorities have recently issued regulations concerning the final disposal of nuclear waste. The principal compliance criterion states that the annual risk of harmful effects must not exceed 10 -6 for a representative individual in the group exposed to the greatest risk. There are also a number of requirements on methodological aspects of the safety assessment as well as on the contents of a safety report. The regulations are reproduced in an Appendix to this report. The primary safety function of the KBS 3 system is to completely isolate the spent nuclear fuel within copper canisters over the entire assessment period, which will be one million years in SR-Can. Should a canister be damaged, the secondary safety function is to retard any releases from the canisters. The main steps of the assessment are the following: 1. Qualitative system description, FEP processing: This step consists of defining a system boundary and of describing the system on a format suitable for the safety assessment. Databases of relevant features, events and processes influencing long-term safety are structured and used as one starting point for the assessment. 2. Initial state descriptions. 3. Process descriptions: In this step all identified processes within the system boundary involved in the long-term evolution of the system are described in detail. 4. Description of boundary conditions: This step is a broad description of the evolution of the boundaries of the system, focussing mainly

Regulatory review of safety cases and safety assessments for near surface

International Nuclear Information System (INIS)

Nys, V.

2003-01-01

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

Safety Auditing and Assessments

Science.gov (United States)

Goodin, James Ronald (Ronnie)

2005-01-01

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

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

The application of fracture mechanics to the safety assessment of transport casks for radioactive materials

International Nuclear Information System (INIS)

Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H.

2004-01-01

BAM is the German responsible authority for the mechanical and thermal design safety assessment of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety proof of package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new ''Guidelines for the Application of Ductile Cast Iron for Transport and Storage Casks for Radioactive Materials''. Based on these guidelines higher stresses than before can become permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof with the help of the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA Advisory Material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will be concluded

The application of fracture mechanics to the safety assessment of transport casks for radioactive material

International Nuclear Information System (INIS)

Zencker, U.; Mueller, K.; Droste, B.; Roedel, R.; Voelzke, H.

2004-01-01

BAM is the responsible authority in Germany for the assessment of the mechanical and thermal design safety of packages for the transport of radioactive materials. The assessment has to cover the brittle fracture safety 'proof of package' for components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new 'Guidelines for the application of ductile cast iron for transport and storage casks for radioactive materials'. Based on these guidelines, higher stresses than before will be permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof using the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA's advisory material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will conclude the paper. (author)

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

International Nuclear Information System (INIS)

Hedin, Allan

2002-12-01

This report is a planning document for the preliminary safety evaluations (PSE) to be carried out at the end of the initial stage of SKBs ongoing site investigations for a deep repository for spent nuclear fuel. The main purposes of the 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 and to provide feed-back to continued site investigations and site specific repository design. The preliminary safety evaluations will be carried out by a safety assessment group, based on a site model, being part of a site description, provided by a site modelling group and a repository layout within that model suggested by a repository engineering group. The site model contains the geometric features of the site as well as properties of the host rock. Several alternative interpretations of the site data will likely be suggested. Also the biosphere is included in the site model. A first task for the PSE will be to compare the rock properties described in the site model to previously established criteria for a suitable host rock. This report gives an example of such a comparison. In order to provide more detailed feedback, a number of thermal, hydrological, mechanical and chemical analyses of the site will also be included in the evaluation. The selection of analyses is derived from the set of geosphere and biosphere analyses preliminarily planned for the comprehensive safety assessment named SR-SITE, which will be based on a complete site investigation. The selection is dictated primarily by the expected feedback to continued site investigations and by the availability of data after the PSE. The repository engineering group will consider several safety related factors in suggesting a repository layout: Thermal calculations will be made to determine a minimum distance between canisters avoiding canister surface temperatures above 100 deg C

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.

Safety and Security of Radioactive Sources: Initiatives of the Forum of Nuclear Regulatory Bodies in Africa (FNRBA)

International Nuclear Information System (INIS)

Severa, R.

2010-01-01

Safety and Security of Radioactive Sources: Initiatives of the Forum of Nuclear Regulatory Bodies in Africa(FNRBA) is a regional organization comprising of nuclear regulatory bodies it’s goals are to promote the establishment of regulatory infrastructure in all countries of the Region to adopt joint action plan for implementation of self-assessment and work with Member States to upgrade their regulatory infrastructures, develop and promote a framework for capacity building in areas of radiation and nuclear safety and security, to create an opportunity for mutual support and coordination of regional initiatives by leveraging the development and utilization of regional and international resources and expertise and to serve as reference body on matters relating to nuclear and radiation safety and security in the Region. Radioactive active sources continue to play an increasingly important role in socio-economic activities on the African continent. There is also an ever increasing need to ensure that radioactive sources are utilized in a safe and secure manner

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)

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

Development and initial validation of an Aviation Safety Climate Scale.

Science.gov (United States)

Evans, Bronwyn; Glendon, A Ian; Creed, Peter A

2007-01-01

A need was identified for a consistent set of safety climate factors to provide a basis for aviation industry benchmarking. Six broad safety climate themes were identified from the literature and consultations with industry safety experts. Items representing each of the themes were prepared and administered to 940 Australian commercial pilots. Data from half of the sample (N=468) were used in an exploratory factor analysis that produced a 3-factor model of Management commitment and communication, Safety training and equipment, and Maintenance. A confirmatory factor analysis on the remaining half of the sample showed the 3-factor model to be an adequate fit to the data. The results of this study have produced a scale of safety climate for aviation that is both reliable and valid. This study developed a tool to assess the level of perceived safety climate, specifically of pilots, but may also, with minor modifications, be used to assess other groups' perceptions of safety climate.

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

Are area-based initiatives able to improve area safety in deprived areas? A quasi-experimental evaluation of the Dutch District Approach.

Science.gov (United States)

Kramer, Daniëlle; Jongeneel-Grimen, Birthe; Stronks, Karien; Droomers, Mariël; Kunst, Anton E

2015-07-28

Numerous area-based initiatives have been implemented in deprived areas across Western-Europe with the aim to improve the socio-economic and environmental conditions in these areas. Only few of these initiatives have been scientifically evaluated for their impact on key social determinants of health, like perceived area safety. Therefore, this study aimed to assess the impact of a Dutch area-based initiative called the District Approach on trends in perceived area safety and underlying problems in deprived target districts. A quasi-experimental design was used. Repeated cross-sectional data on perceived area safety and underlying problems were obtained from the National Safety Monitor (2005-2008) and its successor the Integrated Safety Monitor (2008-2011). Study population consisted of 133,522 Dutch adults, including 3,595 adults from target districts. Multilevel logistic regression analyses were performed to assess trends in self-reported general safety, physical order, social order, and non-victimization before and after the start of the District Approach mid-2008. Trends in target districts were compared with trends in various control groups. Residents of target districts felt less safe, perceived less physical and social order, and were victimized more often than adults elsewhere in the Netherlands. For non-victimization, target districts showed a somewhat more positive change in trend after the start of the District Approach than the rest of the Netherlands or other deprived districts. Differences were only statistically significant in women, older adults, and lower educated adults. For general safety, physical order, and social order, there were no differences in trend change between target districts and control groups. Results suggest that the District Approach has been unable to improve perceptions of area safety and disorder in deprived areas, but that it did result in declining victimization rates.

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

Energy Technology Data Exchange (ETDEWEB)

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

1983-08-01

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

Probabilistic safety assessment of the nuclear facilities in Cuba

International Nuclear Information System (INIS)

Rivero O, J.J.; Salomon L, J.

1991-01-01

During 1986-1990 basis were established for further developing probabilistic safety assessment (PSA) of Juragua NPP. A team work was consolidated and carried out the preliminary studies of the small break LOCA initiating event. A significant achievement was the creation of the ANCON code, which allows the evaluation of complex fault trees in personal computers, and has been applied in PSA modelling, and specialist qualification. The paper describes the main results and future activities in this field. (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

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

Reactor Safety Assessment System

International Nuclear Information System (INIS)

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

1987-01-01

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

Reactor safety assessment system

International Nuclear Information System (INIS)

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

1987-01-01

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

Independent assessment for new nuclear reactor safety

Directory of Open Access Journals (Sweden)

D'Auria Francesco

2017-01-01

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

Independent assessment for new nuclear reactor safety

International Nuclear Information System (INIS)

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

2017-01-01

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

Assessing medical students' perceptions of patient safety: the medical student safety attitudes and professionalism survey.

Science.gov (United States)

Liao, Joshua M; Etchegaray, Jason M; Williams, S Tyler; Berger, David H; Bell, Sigall K; Thomas, Eric J

2014-02-01

To develop and test the psychometric properties of a survey to measure students' perceptions about patient safety as observed on clinical rotations. In 2012, the authors surveyed 367 graduating fourth-year medical students at three U.S. MD-granting medical schools. They assessed the survey's reliability and construct and concurrent validity. They examined correlations between students' perceptions of organizational cultural factors, organizational patient safety measures, and students' intended safety behaviors. They also calculated percent positive scores for cultural factors. Two hundred twenty-eight students (62%) responded. Analyses identified five cultural factors (teamwork culture, safety culture, error disclosure culture, experiences with professionalism, and comfort expressing professional concerns) that had construct validity, concurrent validity, and good reliability (Cronbach alphas > 0.70). Across schools, percent positive scores for safety culture ranged from 28% (95% confidence interval [CI], 13%-43%) to 64% (30%-98%), while those for teamwork culture ranged from 47% (32%-62%) to 74% (66%-81%). They were low for error disclosure culture (range: 10% [0%-20%] to 27% [20%-35%]), experiences with professionalism (range: 7% [0%-15%] to 23% [16%-30%]), and comfort expressing professional concerns (range: 17% [5%-29%] to 38% [8%-69%]). Each cultural factor correlated positively with perceptions of overall patient safety as observed in clinical rotations (r = 0.37-0.69, P safety behavioral intent item. This study provided initial evidence for the survey's reliability and validity and illustrated its applicability for determining whether students' clinical experiences exemplify positive patient safety environments.

Probabilistic safety assessment - regulatory perspective

International Nuclear Information System (INIS)

Solanki, R.B.; Paul, U.K.; Hajra, P.; Agarwal, S.K.

2002-01-01

Full text: Nuclear power plants (NPPs) have been designed, constructed and operated mainly based on deterministic safety analysis philosophy. In this approach, a substantial amount of safety margin is incorporated in the design and operational requirements. Additional margin is incorporated by applying the highest quality engineering codes, standards and practices, and the concept of defence-in-depth in design and operating procedures, by including conservative assumptions and acceptance criteria in plant response analysis of postulated initiating events (PIEs). However, as the probabilistic approach has been improved and refined over the years, it is possible for the designer, operator and regulator to get a more detailed and realistic picture of the safety importance of plant design features, operating procedures and operational practices by using probabilistic safety assessment (PSA) along with the deterministic methodology. At present, many countries including USA, UK and France are using PSA insights in their decision making along with deterministic basis. India has also made substantial progress in the development of methods for carrying out PSA. However, consensus on the use of PSA in regulatory decision-making has not been achieved yet. This paper emphasises on the requirements (e.g.,level of details, key modelling assumptions, data, modelling aspects, success criteria, sensitivity and uncertainty analysis) for improving the quality and consistency in performance and use of PSA that can facilitate meaningful use of the PSA insights in the regulatory decision-making in India. This paper also provides relevant information on international scenario and various application areas of PSA along with progress made in India. The PSA perspective presented in this paper may help in achieving consensus on the use of PSA for regulatory / utility decision-making in design and operation of NPPs

Probabilistic safety assessment of Tehran Research Reactor using systems analysis programs for hands-on integrated reliability evaluations

International Nuclear Information System (INIS)

Hosseini, M.H.; Nematollahi, M.R.; Sepanloo, K.

2004-01-01

Probabilistic safety assessment application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this document the application of the probabilistic safety assessment to the Tehran Research Reactor is presented. The level 1 practicabilities safety assessment application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantifications, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using systems analysis programs for hands-on integrated reliability evaluations software

The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine

International Nuclear Information System (INIS)

Kot, C.

1999-01-01

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments

AGR core safety assessment methodologies

International Nuclear Information System (INIS)

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

1996-01-01

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

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)

Assessment of radiation safety in well logging

International Nuclear Information System (INIS)

Alles, A.; Pérez, Y.; Duménigo, C.

2015-01-01

Radiation safety assessments required by current regulations are a means to verify compliance with the requirements. Different methods have been used for this purpose. In the paper the results of applying the method of risk matrices, applied for the first time in the practice of well logging are exposed. For each initiating event frequency of occurrence, the severity of its consequences and the probability of failure of the barriers identified were evaluated. Starting from these assumptions, the risk associated is determined for each of the identified accident sequences, using for this the SEVRRA code 'Risk Assessment System', originally designed for use in radiotherapy. As an result sequences increased risk associated with the practice of well logging were identified, which is the starting point for the further implementation of a coherent program of dose optimization in practice. [es

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan [and others

2004-08-01

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

Interim main report of the safety assessment SR-Can

International Nuclear Information System (INIS)

Hedin, Allan

2004-08-01

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

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

Danish initiatives to improve the safety of meat products

DEFF Research Database (Denmark)

Wegener, Henrik Caspar

2010-01-01

and Campylobacter, and to a lesser extent Yersinia, Escherichia coli O157 and Listeria. Danish initiatives to improve the safety of meat products have focused on the entire production chain from the farm to the consumer, with a special emphasis on the pre-harvest stage of production. The control of bacterial......During the last two decades the major food safety problems in Denmark, as determined by the number of human patients, has been associated with bacterial infections stemming from meat products and eggs. The bacterial pathogens causing the majority of human infections has been Salmonella...

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

1999-05-15

The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

Procedures for conducting probabilistic safety assessments of nuclear power plants (Level 1)

International Nuclear Information System (INIS)

1992-01-01

This report provides guidance for conducting a Level 1 of probabilistic safety assessment (PSA), that is a PSA concerned with events leading to core damage. The scope of this report is confined to internal initiating events (excluding internal fires and floods). A particular aim is to promote a standardized framework, terminology and form of documentation for PSAs so as to facilitate external review of the results of such studies. The report is divided into the following major sections: management and organization; identification of sources of radioactive releases and accident initiators; accident sequence modelling; data assessment and parameter estimation; accident sequence quantification; documentation of the analysis: display and interpretation of result. 45 refs, 7 figs, 23 tabs

Pushing and pulling: an assessment tool for occupational health and safety practitioners.

Science.gov (United States)

Lind, Carl Mikael

2018-03-01

A tool has been developed for supporting practitioners when assessing manual pushing and pulling operations based on an initiative by two global companies in the manufacturing industry. The aim of the tool is to support occupational health and safety practitioners in risk assessment and risk management of pushing and pulling operations in the manufacturing and logistics industries. The tool is based on a nine-multiplier equation that includes a wide range of factors affecting an operator's health risk and capacity in pushing and pulling. These multipliers are based on psychophysical, physiological and biomechanical studies in combination with judgments from an expert group consisting of senior researchers and ergonomists. In order to consider usability, more than 50 occupational health and safety practitioners (e.g., ergonomists, managers, safety representatives and production personnel) participated in the development of the tool. An evaluation by 22 ergonomists supports that the push/pull tool is user friendly in general.

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)

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

International Nuclear Information System (INIS)

2015-08-01

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

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

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

Uncertainty and sensitivity analysis on probabilistic safety assessment of an experimental facility

International Nuclear Information System (INIS)

Burgazzi, L.

2000-01-01

The aim of this work is to perform an uncertainty and sensitivity analysis on the probabilistic safety assessment of the International Fusion Materials Irradiation Facility (IFMIF), in order to assess the effect on the final risk values of the uncertainties associated with the generic data used for the initiating events and component reliability and to identify the key quantities contributing to this uncertainty. The analysis is conducted on the expected frequency calculated for the accident sequences, defined through the event tree (ET) modeling. This is in order to increment credit to the ET model quantification, to calculate frequency distributions for the occurrence of events and, consequently, to assess if sequences have been correctly selected on the probability standpoint and finally to verify the fulfillment of the safety conditions. Uncertainty and sensitivity analysis are performed using respectively Monte Carlo sampling and an importance parameter technique. (author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Safety 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)

Non-animal approaches for consumer safety risk assessments: Unilever's scientific research programme.

Science.gov (United States)

Carmichael, Paul; Davies, Michael; Dent, Matt; Fentem, Julia; Fletcher, Samantha; Gilmour, Nicola; MacKay, Cameron; Maxwell, Gavin; Merolla, Leona; Pease, Camilla; Reynolds, Fiona; Westmoreland, Carl

2009-12-01

Non-animal based approaches to risk assessment are now routinely used for assuring consumer safety for some endpoints (such as skin irritation) following considerable investment in developing and applying new methods over the past 20 years. Unilever's research programme into non-animal approaches for safety assessment is currently focused on the application of new technologies to risk assessments in the areas of skin allergy, cancer and general toxicity (including inhalation toxicity). In all of these areas, a long-term investment is essential to increase the scientific understanding of the underlying biological and chemical processes that we believe will ultimately form a sound basis for novel risk assessment approaches. Our research programme in these priority areas consists of in-house research as well as Unilever-sponsored academic research, involvement with EU-funded projects (e.g. Sens-it-iv, carcinoGENOMICS), participation in cross-industry collaborative research (e.g. COLIPA, EPAA) and ongoing involvement with other scientific initiatives on non-animal approaches to risk assessment (e.g. UK NC3Rs, US 'Human Toxicology Project' consortium). 2009 FRAME.

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

Savannah River Site management response plan for chemical safety vulnerability field assessment. Revision 1

International Nuclear Information System (INIS)

Kahal, E.J.; Murphy, S.L.; Salaymeh, S.R.

1994-09-01

As part of the U.S. Department of Energy's (DOE) initiative to identify potential chemical safety vulnerabilities in the DOE complex, the Chemical Safety Vulnerability Core Working Group issued a field verification assessment report. While the report concluded that Savannah River Site (SRS) is moving in a positive direction, the report also identified five chemical safety vulnerabilities with broad programmatic impact that are not easily nor quickly remedied. The May 1994 SRS Management Response Plan addressed the five SRS vulnerabilities identified in the field assessment report. The SRS response plan listed observations supporting the vulnerabilities and any actions taken or planned toward resolution. Many of the observations were resolved by simple explanations, such as the existence of implementation plans for Safety Analysis Report updates. Recognizing that correcting individual observations does not suffice in remedying the vulnerabilities, a task team was assembled to address the broader programmatic issues and to recommend corrective actions

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan (ed.) [and others

2004-08-01

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

The Nirex safety assessment research programme

International Nuclear Information System (INIS)

Cooper, M.J.

1988-07-01

This report describes progress on the Nirex Safety Assessment Research Programme in 1987/88. The programme is concerned with research into the disposal of low-level waste (LLW) and intermediate-level waste (ILW) into underground repositories. At the beginning of 1987/88 a range of techniques for measuring and modelling far-field phenomena were being applied to near-surface disposal of low-level waste in clay. However, during the year the far-field studies were redirected to consider generic geological materials of interest for deep disposal of low and intermediate-level waste, which is now the preferred option in the UK. A substantial part of the programme is concerned with the effectiveness of near-field barriers to water-borne leakage of radionuclides from cementitious repositories. Considerable progress has been made in quantifying this and laying the foundations for robust and reliable radiological assessments to be made with appropriate models. New projects have also been initiated to study the evolution and migration of gases from an underground repository and to consider the contribution of the biosphere to the retardation of radionuclides. (author)

Safety assessment of foods derived from genetically modified crops

NARCIS (Netherlands)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

2003-01-01

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

Assessment of safety culture: Changing regulatory approach in Hungary

International Nuclear Information System (INIS)

Ronaky, Jozsef; Toth, Andras

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

The radiation safety self-assessment program of Ontario Hydro

International Nuclear Information System (INIS)

Armitage, G.; Chase, W.J.

1987-01-01

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

Initiating events in the safety probabilistic analysis of nuclear power plants

International Nuclear Information System (INIS)

Stasiulevicius, R.

1989-01-01

The importance of the initiating event in the probabilistic safety analysis of nuclear power plants are discussed and the basic procedures necessary for preparing reports, quantification and grouping of the events are described. The examples of initiating events with its occurence medium frequency, included those calculated for OCONEE reactor and Angra-1 reactor are presented. (E.G.)

Probabilistic safety assessment for seismic events

International Nuclear Information System (INIS)

1993-10-01

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

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

International Nuclear Information System (INIS)

2008-12-01

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

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

Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

International Nuclear Information System (INIS)

Torabi, T.; Bettig, R.; Iliescu, P.; Robinson, J.; Santamaura, P.; Skorupska, B.; Tyagi, A.K.; Vencel, I.

2013-01-01

The Enhanced CANDU 6(EC6) is a 700 MWe reactor, which has evolved from the well-established CANDU line of reactors, which are heavy-water moderated, and heavy-water cooled horizontal pressure tube reactors, using natural uranium fuel. The EC6 design retains the generic CANDU design features, while incorporating innovations and state-of-the-art technologies to ensure competitiveness with other design with respect to operation, performance and economics. A design assist probabilistic safety assessment (PSA) was conducted during the design change phase of the project. The purpose of the assessment was to assess internal events during at-power operation and identify the design improvements and additional features needed to comply with the latest regulatory requirements in Canada and compete with other reactor designs, internationally. The PSA results show that the EC6 plant response to the postulated initiating events is well balanced, and the design meets its safety objectives. This paper summarizes the results and insights gained during the development of the PSA models for at-power internal events. (author)

OECD/NEA WGFCS Workshop: Safety Assessment of Fuel Cycle Facilities - Regulatory Approaches and Industry Perspectives

International Nuclear Information System (INIS)

2013-01-01

Nuclear fuel is produced, processed, and stored mainly in industrial-scale facilities. Uranium ores are processed and refined to produce a pure uranium salt stream, Uranium is converted and enriched, nuclear fuel is fabricated (U fuel and U/Pu fuel for the closed cycle option); and spent fuel is stored and reprocessed in some countries (close cycle option). Facilities dedicated to the research and development of new fuel or new processes are also considered as Fuel Cycle Facilities. The safety assessment of nuclear facilities has often been led by the methodology and techniques initially developed for Nuclear Power Plants. As FCFs cover a wide diversity of installations the various approaches of national regulators, and their technical support organizations, for the Safety Assessment of Fuel Cycle Facilities are also diverse, as are the approaches by their industries in providing safety justifications for their facilities. The objective of the Working Group on Fuel Cycle Safety is to advance the understanding for both regulators and operators of relevant aspects of nuclear fuel cycle safety in member countries. A large amount of experience is available in safety assessment of FCFs, which should be shared to develop ideas in this field. To contribute to this task, the Workshop on 'Safety Assessment of Fuel Cycle Facilities - Regulatory Approaches and Industry Perspectives' was held in Toronto, on 27 - 29 September 2011. The workshop was hosted by Canadian Nuclear Safety Commission. The current proceedings provide summary of the results of the workshop with the text of the papers given and presentations made

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

Use of operational experience in fire safety assessment of nuclear power plants

International Nuclear Information System (INIS)

2000-01-01

Fire hazard has been identified as a major contributor to a plant's operational risk and the international nuclear power industry has been studying and developing tools for defending against this hazard. Considerable progress in design and regulatory requirements for fire safety, in fire protection technology and in related analytical techniques has been made in the past two decades. Substantial efforts have been undertaken worldwide to implement these advances in the interest of improving fire safety both at new and existing nuclear power plants. To assist in these efforts, the IAEA initiated a programme on fire safety that was intended to provide assistance to Member States in improving fire safety in nuclear power plants. In order to achieve this general objective, the IAEA programme aimed at the development of guidelines and good practices, the promotion of advanced fire safety assessment techniques, the exchange of state of the art information between practitioners and the provision of engineering safety advisory services and training in the implementation of internationally accepted practices. During the period 1993-1994, the IAEA activities related to fire safety concentrated on the development of guidelines and good practice documents related to fire safety and fire protection of operating plants. One of the first tasks was the development of a Safety Guide that formulates specific requirements with regard to the fire safety of operating nuclear power plants. Several documents, which provide advice on fire safety inspection, were developed to assist in its implementation. In the period 1995-1996, the programme focused on the preparation of guidelines for the systematic analysis of fire safety at nuclear power plants (NPPs). The IAEA programme on fire safety for 1997-1998 includes tasks aimed at promoting systematic assessment of fire safety related occurrences and dissemination of essential insights from this assessment. One of the topics addressed is the

Types of safety assessments of near surface repository for radioactive waste

International Nuclear Information System (INIS)

Mateeva, M.

2004-01-01

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

Data used for safety assessment of reprocessing facilities

International Nuclear Information System (INIS)

Nomura, Yasushi; Suzuki, Atsuyuki; Kanagawa, Akira

1990-08-01

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

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

International Nuclear Information System (INIS)

Hedin, A.

2008-01-01

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

Safety and reliability assessment

International Nuclear Information System (INIS)

1979-01-01

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

Effects of Implemented Initiatives on Patient Safety Culture in Fateme Al-zahra Hospital in Najafabad

Directory of Open Access Journals (Sweden)

Ahmadreza Izadi

2015-01-01

Full Text Available Introduction: Patient safety improvement requires ongoing culture. This cultural change is the most important challenge that managers are faced with in creation of a safe system. This study aims to show the results of initiatives to improvement in patient safety culture in Fateme Al-zahra hospital. Method: In the quasi-experimental research, patient safety culture was measured using the Persian questionnaire on adaptation of the hospital survey on patient safety culture in 12 dimensions. The research was conducted before (January 2010 and after (September 2012 the improvement initiatives. In this study, all units were determined and no sampling method was used. Reliability of the questionnaire was tested by Alpha Chronbakh (0.83. Data were analyzed using descriptive statistics indices and Independent T-Test by SPSS Software (version 18. Results: 350 questionnaires were distributed in each phaseand overall response rate was 58 and 56 percent, respectively. According to Independent T-test, Management expectations and actions, Organizational learning, Management support, Feedback and communication about error, Communication openness, Overall Perceptions of Safety, Non-punitive Response to Error, Frequency of Event Reporting, and Patient safety culture showed significant differences (P-value0.05. The mean score of Patient safety culture was 2.27 (from 5 and it was increased to 2.46 after initiatives that showed a significant difference (P-value<0.05. Conclusion: Although, improvement in patient safety culture needs teamwork and continuous attempts, the study showed that initiatives implemented in the case hospital had been effective in some dimensions. However, Teamwork within hospital units, Teamwork across units, Hospital handoffs and transitions, and Staffing dimensions were recognized for further intervention. Hospital could improve the patient safety culture with planning and measures in these dimensions.

EFFICIENCY AND SAFETY ASSESSMENT OF GENERIC ATORVASTATINE IN PATIENTS WITH HYPERLIPIDEMIA

Directory of Open Access Journals (Sweden)

J. E. Semyonova

2015-12-01

Full Text Available Aim. To assess in a short-term study efficiency and safety of hypolipidemic therapy with atorvastatine generic, Tulip, in comparison with simvastatine generic, Vasilip, in hyperlipidemic patients.Material and methods. Open, randomized, comparative, cross over study included 87 patients with hyperlipidemia, who didn’t receive hypolipidemic drugs within 6 weeks, or followed hypolipidemic diet for 4 weeks. Each patient received therapy with one of the studied drugs within 6 weeks. Then after 4-week wash-out period the second therapy with the other drug was held. Consequence of courses with each drug was set by randomization. Initial dose of both drugs was 10 mg daily. Dose was adjusted after 3 weeks. The dose was increased to 20 mg daily if cholesterol of low density lipoproteid (CLDL hadn’t reached target level (< 115 mg/dl. of Treatment safety was assessed on the basis of clinical data, hepatic enzymes activity and creatine phosphokinase levels.Results. It is shown, that to reach target figures of plasma lipid spectrum, Vasilip dose was increased significantly more often, than Tulip dose. Average Tulip dose after titration was 14,8 mg daily, Vasilip dose – 15,6 mg daily. Patients with initially higher level of triglycerides (TG > 170 mg/dl after 6 weeks with Tulip treatment showed TG reduction by 38% and with Vasilip treatment – by 20%. Both drugs showed good tolerance, no significant differences in number of side-effects were observed.Conclusion. 6-week treatment with atorvastatine generic Tulip shows significant hypolipidemic effect, which appears in significant reduction of CLDL, total cholesterol, TG compared to the initial levels. Degree of total cholesterol reduction is significantly higher with Tulip treatment compared to Vasilip treatment. Analyses shown that target levels of the assessed figures were reached in more patients, treated with Tulip. Side-effects in Tulip treatment were not severe.

EFFICIENCY AND SAFETY ASSESSMENT OF GENERIC ATORVASTATINE IN PATIENTS WITH HYPERLIPIDEMIA

Directory of Open Access Journals (Sweden)

J. E. Semyonova

2005-01-01

Full Text Available Aim. To assess in a short-term study efficiency and safety of hypolipidemic therapy with atorvastatine generic, Tulip, in comparison with simvastatine generic, Vasilip, in hyperlipidemic patients.Material and methods. Open, randomized, comparative, cross over study included 87 patients with hyperlipidemia, who didn’t receive hypolipidemic drugs within 6 weeks, or followed hypolipidemic diet for 4 weeks. Each patient received therapy with one of the studied drugs within 6 weeks. Then after 4-week wash-out period the second therapy with the other drug was held. Consequence of courses with each drug was set by randomization. Initial dose of both drugs was 10 mg daily. Dose was adjusted after 3 weeks. The dose was increased to 20 mg daily if cholesterol of low density lipoproteid (CLDL hadn’t reached target level (< 115 mg/dl. of Treatment safety was assessed on the basis of clinical data, hepatic enzymes activity and creatine phosphokinase levels.Results. It is shown, that to reach target figures of plasma lipid spectrum, Vasilip dose was increased significantly more often, than Tulip dose. Average Tulip dose after titration was 14,8 mg daily, Vasilip dose – 15,6 mg daily. Patients with initially higher level of triglycerides (TG > 170 mg/dl after 6 weeks with Tulip treatment showed TG reduction by 38% and with Vasilip treatment – by 20%. Both drugs showed good tolerance, no significant differences in number of side-effects were observed.Conclusion. 6-week treatment with atorvastatine generic Tulip shows significant hypolipidemic effect, which appears in significant reduction of CLDL, total cholesterol, TG compared to the initial levels. Degree of total cholesterol reduction is significantly higher with Tulip treatment compared to Vasilip treatment. Analyses shown that target levels of the assessed figures were reached in more patients, treated with Tulip. Side-effects in Tulip treatment were not severe.

Assessment of safety culture at INPP

International Nuclear Information System (INIS)

Lesin, S.

2002-01-01

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

Safety assessment for radioactive waste disposal facility

International Nuclear Information System (INIS)

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

2008-08-01

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

Probabilistic safety assessment of the radiotherapy treatment with a linear accelerator for medical use

International Nuclear Information System (INIS)

Vilaragut Llanes, Juan Jose; Ferro Fernandez, Ruben; Rodriguez MartI, Manuel; Ramirez, Maria Luisa; Perez Mulas, Arturo; Barrientos Montero, Marta; Ortiz Lopez, Pedro; Somoano, Fernando; Delgado RodrIguez, Jose Miguel; Papadopulos, Susana B.; Pereira Jr, Pedro Paulo; Lopez Morones, Ramon; Larrinaga Cortina, Eduardo; Rivero Oliva, Jose de Jesus; Alemanny, Jorge

2010-01-01

This paper presents the results of the Probabilistic Safety Assessment to the radiotherapy treatment with an Electron Linear Accelerator for Medical Use, which was conducted in the framework of the Iberian-American Forum of Radiological and Nuclear Regulatory Agencies. Potential accidental exposures during the treatment of patients, workers and members of the public were assessed, although the study was mainly focused on patients. The methodology of failure modes and effects analysis was used to define accident initiating events and methods of event tree and fault tree analysis to determine the accident sequences that may occur. After quantifying the frequency of occurrence of the accident sequences, an important analysis was carried out in order to determine the most significant events from the point of view of safety. The major contributors to risk were identified as well as the most appropriate safety recommendations to reduce it. (author)

First safety assessment objectives and content of the 2001 report

International Nuclear Information System (INIS)

Franco, Michel de

2002-01-01

Michel de Franco (ANDRA, France) described plans to report its first safety assessment of facility designs for disposal of high and intermediate level waste and spent fuel. This assessment will be the forerunner of a more detailed assessment that is required to be presented to the French government in 2005 and is intended to facilitate the formalization and testing of the assessment methodology intended to be used in the 2005 assessment report. The report will include information about the waste inventory, the materials used for the engineered barriers and current understanding of the geology and surface environment at the Bure site in eastern France. It will also describe the preliminary design concepts and the phenomena defining the evolution of the repository in different time frames as well as presenting the results of the initial performance assessment of the repository. The report will also include an analysis of the implications of the requirement for reversibility, taken to mean that each repository development step can be reversed

Integrated Plant Safety Assessment, Systematic Evaluation Program: Yankee Nuclear Power Station (Docket No. 50-29)

International Nuclear Information System (INIS)

1987-10-01

The US Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0825), under the scope of the Systematic Evaluation Program (SEP), for Yankee Atomic Electric Company's Yankee Nuclear Power Station located in Rowe, Massachusetts. The SEP was initiated by the NRC to review the design of older operating nuclear power plants to reconfirm and document their safety. This report documents the review completed under the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations after the Final IPSAR for the Yankee plant was issued. The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when Yankee was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. 2 tabs

Preliminary safety assessment of the WIPP facility

International Nuclear Information System (INIS)

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

1979-01-01

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

LNG Safety Assessment Evaluation Methods

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-01

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

Promoting and assessment of safety culture within regulatory body

International Nuclear Information System (INIS)

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

2015-01-01

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

Probabilistic safety assessment for research reactors

International Nuclear Information System (INIS)

1986-12-01

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

Quantification of the impact of multifaceted initiatives intended to improve operational efficiency and the safety culture: a case study from an academic medical center radiation oncology department.

Science.gov (United States)

Chera, Bhishamjit S; Mazur, Lukasz; Jackson, Marianne; Taylor, Kinely; Mosaly, Prithima; Chang, Sha; Deschesne, Kathy; LaChapelle, Dana; Hoyle, Lesley; Saponaro, Patricia; Rockwell, John; Adams, Robert; Marks, Lawrence B

2014-01-01

We have systematically been incorporating several operational efficiency and safety initiatives into our academic radiation oncology clinic. We herein quantify the impact of these initiatives on prospectively collected, clinically meaningful, metrics. The data from 5 quality improvement initiatives, each focused on a specific safety/process concern in our clinic, are presented. Data was collected prospectively: operational metrics recorded before and after implementation of the initiative were compared using statistical analysis. Results from the Agency for Health Care Research and Quality (AHRQ) patient safety culture surveys administered during and after many of these initiatives were similarly compared. (1) Workload levels for nurses assisting with brachytherapy were high (National Aeronautics and Space Administration Task Load Index (NASA-TLX) scores >55-60, suggesting, "overwork"). Changes in work flow and procedure room layout reduced workload to more acceptable levels (NASA-TLX 50% to <10%; P < .01). To assess the overall changes in "patient safety culture," we conducted a pre- and postanalysis using the AHRQ survey. Improvements in all measured dimensions were noted. Quality improvement initiatives can be successfully implemented in an academic radiation oncology department to yield measurable improvements in operations resulting in improvement in patient safety culture. Copyright © 2014 American Society for Radiation Oncology. Published by Elsevier Inc. All rights reserved.

Safety assessment of radioactive wastes storage 'Mironova Gora'

International Nuclear Information System (INIS)

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

2000-01-01

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

Probabilistic safety assessment activities at Ignalina NPP

International Nuclear Information System (INIS)

Bagdonas, A.

1999-01-01

The Barselina Project was initiated in the summer 1991. The project was a multilateral co-operation between Lithuania, Russia and Sweden up until phase 3, and phase 4 has been performed as a bilateral between Lithuania and Sweden. The long-range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. During phase 3, from 1993 to 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. During phase 4, from 1994 to 1996, the PSA was further developed, taking into account plant changes, improved modelling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The model reflected the plant status before the outage 1996. During phase 4+, 1998 to 1999 the PSA model was upgraded taking into account the newest plant modifications. The new PSA model of CPS/AZRT was developed. Modelling was based on the Single Failure Analysis

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

Kozloduy nuclear power plant. Units 1-4. Status of safety assessment activities. Rev. 2

International Nuclear Information System (INIS)

1999-01-01

This paper presents the results of the status of safety assessment activities carried out by the Kozloduy Nuclear Power Plant (KNPP) in order to evaluate the current status of the safety of its reactor units 1-4. The steam supply system of this units is based of the reactor WWER-440/ B-230, which is a PWR of Russian design developed according to the safety standards in force in USSR in late 60-s. Now a days 10 reactor units of this type are in operation in four NPPs. Despite of efforts of the different plants to implement safety improvements measures during first 10-15 years of operation of this type of reactor its major safety problems were not eliminated and were a subject of international concern. The systematic evaluation of the deficiencies of the original design of this type of reactors have been initiated by IAEA in the beginning of 1990 and brought to developing a comprehensive list of safety problems which required urgent implementation of safety measures in all plants. To solve this problems in 1991 KNPP initiated implementation of so called 'short term' safety improvement program, developed with the help of WANO under agreement with Bulgarian Nuclear Safety Authority (BNSA) and consortium RISKAUDIT. The program was based on a stage approach and was foreseen to be implemented by tree stages in very tight time schedule in order to achieve significant and rapid improvements of the level of safety in operation of the units. The Short Tenn Program was implemented between the years 1991 and 1997 thanks of the strong safety commitment of NEK and KNPP staff and the broad international cooperation and financial support. Important part of resources were supplied under PHARE program of CEC, EBRD grant agreement and EDF support. The plant current safety level analysis has been performed using IAEA analytical methodology according to 50-SG-O12 standard 'Periodic safety review of operational nuclear power plants'. The approach and criteria for acceptable safety level

A probabilistic safety assessment of in-pile test loop in HWRR

International Nuclear Information System (INIS)

Cao Xuewu; Li Zhaohuan

1991-07-01

The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

Self-assessment of operational safety for nuclear power plants

International Nuclear Information System (INIS)

1999-12-01

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

Analysis of truncation limit in probabilistic safety assessment

International Nuclear Information System (INIS)

Cepin, Marko

2005-01-01

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

Risk assessment of safety violations for coal mines

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-15

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

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

OpenAIRE

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

2016-01-01

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

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

International Nuclear Information System (INIS)

1994-01-01

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

A review of the probabilistic safety assessment application to the TR-2 research reactor

International Nuclear Information System (INIS)

Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

2001-01-01

A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

Companies' opinions and acceptance of global food safety initiative benchmarks after implementation.

Science.gov (United States)

Crandall, Phil; Van Loo, Ellen J; O'Bryan, Corliss A; Mauromoustakos, Andy; Yiannas, Frank; Dyenson, Natalie; Berdnik, Irina

2012-09-01

International attention has been focused on minimizing costs that may unnecessarily raise food prices. One important aspect to consider is the redundant and overlapping costs of food safety audits. The Global Food Safety Initiative (GFSI) has devised benchmarked schemes based on existing international food safety standards for use as a unifying standard accepted by many retailers. The present study was conducted to evaluate the impact of the decision made by Walmart Stores (Bentonville, AR) to require their suppliers to become GFSI compliant. An online survey of 174 retail suppliers was conducted to assess food suppliers' opinions of this requirement and the benefits suppliers realized when they transitioned from their previous food safety systems. The most common reason for becoming GFSI compliant was to meet customers' requirements; thus, supplier implementation of the GFSI standards was not entirely voluntary. Other reasons given for compliance were enhancing food safety and remaining competitive. About 54 % of food processing plants using GFSI benchmarked schemes followed the guidelines of Safe Quality Food 2000 and 37 % followed those of the British Retail Consortium. At the supplier level, 58 % followed Safe Quality Food 2000 and 31 % followed the British Retail Consortium. Respondents reported that the certification process took about 10 months. The most common reason for selecting a certain GFSI benchmarked scheme was because it was widely accepted by customers (retailers). Four other common reasons were (i) the standard has a good reputation in the industry, (ii) the standard was recommended by others, (iii) the standard is most often used in the industry, and (iv) the standard was required by one of their customers. Most suppliers agreed that increased safety of their products was required to comply with GFSI benchmarked schemes. They also agreed that the GFSI required a more carefully documented food safety management system, which often required

Safety Assessment of Polyether Lanolins as Used in Cosmetics.

Science.gov (United States)

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

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

Probabilistic safety assessment as a standpoint for decision making

International Nuclear Information System (INIS)

Cepin, M.

2001-01-01

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

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

International Nuclear Information System (INIS)

Alujevic, Luka

2014-01-01

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

Development of Risk Assessment Matrix for NASA Engineering and Safety Center

Science.gov (United States)

Malone, Roy W., Jr.; Moses, Kelly

2004-01-01

This paper describes a study, which had as its principal goal the development of a sufficiently detailed 5 x 5 Risk Matrix Scorecard. The purpose of this scorecard is to outline the criteria by which technical issues can be qualitatively and initially prioritized. The tool using this score card has been proposed to be one of the information resources the NASA Engineering and Safety Center (NESC) takes into consideration when making decisions with respect to incoming information on safety concerns across the entire NASA agency. The contents of this paper discuss in detail each element of the risk matrix scorecard, definitions for those elements and the rationale behind the development of those definitions. This scorecard development was performed in parallel with the tailoring of the existing Futron Corporation Integrated Risk Management Application (IRMA) software tool. IRMA was tailored to fit NESC needs for evaluating incoming safety concerns and was renamed NESC Assessment Risk Management Application (NAFMA) which is still in developmental phase.

Safety factors for neutron fluences in NPP safety assessment

International Nuclear Information System (INIS)

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

2016-01-01

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

Ensuring the quality of occupational safety risk assessment.

Science.gov (United States)

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

2013-03-01

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

SCART guidelines. Reference report for IAEA Safety Culture Assessment Review Team (SCART)

International Nuclear Information System (INIS)

2008-01-01

The IAEA Director General stressed the role of safety culture in his concluding remarks at the Meeting of the Contracting Parties to the Convention on Nuclear Safety in 2002: 'As we have learned in other areas, it is not enough simply to have a structure; it is not enough to say that we have the necessary laws and the appropriate regulatory bodies. All these are important, but equally important is that we have in place a safety culture that gives effect to the structure that we have developed. To me, effectiveness and transparency are keys. So, it is an issue which I am pleased to see, you are giving the attention it deserves and we will continue to work with you in clarifying, developing and applying safety culture through our programmes and through our technical cooperation activities.' The concept of safety culture was initially developed by the International Nuclear Safety Advisory Group (INSAG) after the Chernobyl accident in 1986. Since then the IAEA's perspective of safety culture has expanded with time as its recognition of the complexities of the concept developed. Safety culture is considered to be specific organizational culture in all types of organizations with activities that give rise to radiation risks. The aim is to make safety culture strong and sustainable, so that safety becomes a primary focus for all activities in such organizations, even for those, which might not look safety-related at first. SCART (Safety Culture Assessment Review Team) is a safety review service, which reflects the expressed interest of Members States for methods and tools for safety culture assessment. It is a replacement for the earlier service ASCOT (Assessment of Safety Culture in Organizations Team). The IAEA Safety Fundamentals, Requirements and Guides (Safety Standards) are the basis for the SCART Safety Review Service. The reports of INSAG, identifying important current nuclear safety issues, serve also as references during a SCART mission. SCART missions are based

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

Current activities and future trends in reliability analysis and probabilistic safety assessment in Hungary

International Nuclear Information System (INIS)

Hollo, E.; Toth, J.

1986-01-01

In Hungary reliability analysis (RA) and probabilistic safety assessment (PSA) of nuclear power plants was initiated 3 years ago. First, computer codes for automatic fault tree analysis (CAT, PREP) and numerical evaluation (REMO, KITT1,2) were adapted. Two main case studies - detailed availability/reliability calculation of diesel sets and analysis of safety systems influencing event sequences induced by large LOCA - were performed. Input failure data were taken from publications, a need for failure and reliability data bank was revealed. Current and future activities involves: setup of national data bank for WWER-440 units; full-scope level-I PSA of PAKS NPP in Hungary; operational safety assessment of particular problems at PAKS NPP. In the present article the state of RA and PSA activities in Hungary, as well as the main objectives of ongoing work are described. A need for international cooperation (for unified data collection of WWER-440 units) and for IAEA support (within Interregional Program INT/9/063) is emphasized. (author)

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

NARCIS (Netherlands)

Karanikas, Nektarios

2016-01-01

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

NPP Krsko periodic safety review. Safety assessment and analyses

International Nuclear Information System (INIS)

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

2002-01-01

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

Safety assessment of a lithium target

International Nuclear Information System (INIS)

Burgazzi, Luciano; Roberta, Ferri; Barbara, Giannone

2006-01-01

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

The role of natural analogues in safety assessment and acceptability

International Nuclear Information System (INIS)

Papp, Toenis

1987-01-01

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

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

International Nuclear Information System (INIS)

Armstrong, D.L.

1994-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Armstrong, D.L.

1994-08-01

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

Safety, feasibility and efficacy of a rapid ART initiation in pregnancy ...

African Journals Online (AJOL)

Safety, feasibility and efficacy of a rapid ART initiation in pregnancy pilot programme in Cape Town, South Africa. S Black, R Zulliger, L Myer, R Marcus, S Jeneker, R Taliep, D Pienaar, R Wood, L-G Bekker ...

RSAS: a Reactor Safety Assessment System

International Nuclear Information System (INIS)

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

1985-01-01

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

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

Science.gov (United States)

Arch, B N; Thurston, M N

2013-06-01

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

Expensive blood safety initiatives may offer less benefit than we think

DEFF Research Database (Denmark)

Kamper-Jørgensen, Mads; Hjalgrim, Henrik; Edgren, Gustaf

2010-01-01

Various blood safety initiatives have ensured a historically low risk of infection transmission through blood transfusion. Although further prevention of infection transmission is possible through, for example, nucleic acid testing and future introduction of pathogen inactivation, such initiative...... are very costly in relation to the benefit they offer. Although estimation of the cost-effectiveness requires detailed information about the survival of transfusion recipients, previous cost-effectiveness analyses have relied on incorrect survival assumptions....

Development and Psychometric Analysis of a Nurses' Attitudes and Skills Safety Scale: Initial Results.

Science.gov (United States)

Armstrong, Gail E; Dietrich, Mary; Norman, Linda; Barnsteiner, Jane; Mion, Lorraine

Health care organizations have incorporated updated safety principles in the analysis of errors and in norms and standards. Yet no research exists that assesses bedside nurses' perceived skills or attitudes toward updated safety concepts. The aims of this study were to develop a scale assessing nurses' perceived skills and attitudes toward updated safety concepts, determine content validity, and examine internal consistency of the scale and subscales. Understanding nurses' perceived skills and attitudes about safety concepts can be used in targeting strategies to enhance their safety practices.

Modifications of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany based upon new version of Emergency Operating Procedures

International Nuclear Information System (INIS)

Aldorf, R.

1997-01-01

In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to reflect on Probabilistic Safety Assessment-1 basis on impact of Emergency Response Guidelines (as one particular event from the list of other modifications) on Plant Safety. Following highlights help to orient the reader in main general aspects, findings and issues of the work that currently continues on. Older results of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany have revealed that human behaviour during accident progression scenarios represent one of the most important aspects in plant safety. Current effort of Nuclear Power Plants Dukovany (Czech Republic) and Bohunice (Slovak Republic) is focussed on development of qualitatively new symptom-based Emergency Operating Procedures called Emergency Response Guidelines Supplier - Westinghouse Energy Systems Europe, Brussels works in cooperation with teams of specialist from both Nuclear Power Plants. In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to prove on Probabilistic Safety Assessment -1 basis an expected - positive impact of Emergency Response Guidelines on Plant Safety, Since this contract is currently still in progress, it is possible to release only preliminary conclusions and observations. Emergency Response Guidelines compare to original Emergency Operating Procedures substantially reduce uncertainty of general human behaviour during plant response to an accident process. It is possible to conclude that from the current scope Probabilistic Safety Assessment Dukovany point of view (until core damage), Emergency Response Guidelines represent adequately wide basis for mitigating any initiating event

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.

Probabilistic safety analysis on an SBWR 72 hours after the initiating event

International Nuclear Information System (INIS)

Dominguez Bautista, M.T.; Peinador Veira, M.

1996-01-01

Passive plants, including SBWRs, are designed to carry out safety functions with passive systems during the first 72 hours after the initiation event with no need for manual actions or external support. After this period, some recovery actions are required to enable the passive systems to continue performing their safety functions. The study was carried out by the INITEC-Empresarios Agrupados Joint Venture within the framework of the international group collaborating with GE on this project. Its purpose has been to assess, by means of probabilistic criteria, the importance to safety of each of these support actions, in order to define possible requirements to be considered in the design in respect of said recovery actions. In brief, the methodology developed for this objective consists of (1) quantifying success event trees from the PSA up to 72 hours, (2) determining the actions required in each sequence to maintain Steady State after 72 hours, (3) identifying available alternative core cooling methods in each sequence, (4) establishing the approximate (order of magnitude) realizability of each alternative method, (5) calculating the frequency of core damage as a function of the failure probability of post-72-hour actions and (6) analysing the importance of post-72-hour actions. The results of this analysis permit the establishment, right from the conceptual design phase, of the requirements that will arise to ensure these actions in the long term, enhancing their reliability and preventing the accident from continuing beyond this period. (Author)

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

Probabilistic safety assessment of Narora Atomic Power Project

International Nuclear Information System (INIS)

Babar, A.K.; Saraf, R.K.; Kakodkar, A.; Sanyasi Rao, V.V.S.

1989-01-01

Various safety studies on Pressurised Water and Boiling Water reactors have been conducted. However, a detailed report on probabilistic safety assessment (PSA) of PHWRs is not available. PSA level I results of the standardised 235 MWe PHWR under construction at Narora are presented herein. Fault Tree analysis of various initiating events (IEs), safety systems has been completed. Event Tree analysis has been performed for all the dominating IEs to identify the accident sequences and a list of the dominating accident sequences is included. Analysis has been carried out using Monte Carlo simulation to propagate the uncertanities in failure rate data. Further uncertainty analysis is extended to obtain distributions for the accident sequences and core damage frequency. Some noteworthy results of the study apart from the various design modifications incorporated during the design phase are: (i) The accident sequences resulting from station blackout are dominant contributors to the core damage frequency. (ii) Class-IV transients, small break LOCA are significant IEs. Main steam line break is likely to induce steam generator tube ruptures. (iii) Moderator circulation, fire fighting system, secondary steam relief are relatively important in core damage frequency reductions. (iv) Under accidental situations human errors are likely to be asociated with valving in shutdown cooling and fire fighting systems. (author). 14 tabs., 14 figs., 15 refs

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

1996-01-01

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

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

International Nuclear Information System (INIS)

Syarip; Hauptmanns, U.

2000-01-01

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

A generalized framework for assessment of safety margins in nuclear power plants

International Nuclear Information System (INIS)

Gavrilas, M.; Youngblood, B.; Prelewicz, D.; Meyer, Jim

2004-01-01

The protection of public health and safety, and the environment from inadvertent releases of radioactive materials from nuclear power plants relies on the implementation of the defense-in-depth strategy. The term defense-in-depth evolved historically, and thus its application has not always been uniform. The use of the term in the context of the U.S. Nuclear Regulatory Commission (NRC) safety philosophy entails the reliance of a nuclear facility on successive compensatory measures in preventing accidents or mitigating damage caused by malfunctions, accidents, or naturally occurring events. The introduction of probabilistic risk analyses with NUREG-74/014 and subsequent evolution in risk assessment techniques, are leading to the implementation of risk informed regulation to ensure the safety of the public and the environment. Risk informed regulation minimizes the likelihood of overlooking potentially significant accident sequences while limiting unnecessary burdens imposed on licensees. The proposed framework merges fundamental elements of safety regulation: defense-in depth, safety margins and probabilistic risk. It formalizes the relationship between probabilistic risk assessment (PRA) methods and data, and deterministic analyses in a manner consistent with NRC's defense-in-depth philosophy. Succinctly put, the likelihood and consequences of accident scenarios are considered simultaneously and quantified by a plant safety metric. The integration of these fundamental elements into a practically applicable safety framework is consistent with the NRC policy statement on use of probabilistic risk assessment methods and the November 2002 Regulatory Guide on risk informed decisions on plant-specific changes to the licensing basis. Safety information resulting from the application of the framework supersedes traditional safety figures of merit. Safety quantifiers, referred herein as safety indices, expand on the qualifier outcomes that currently accompany fault tree

Development and Psychometric Analysis of a Nurses’ Attitudes and Skills Safety Scale: Initial Results

Science.gov (United States)

Armstrong, Gail E.; Dietrich, Mary; Norman, Linda; Barnsteiner, Jane; Mion, Lorraine

2016-01-01

Health care organizations have incorporated updated safety principles in the analysis of errors and in norms and standards. Yet no research exists that assesses bedside nurses’ perceived skills or attitudes toward updated safety concepts. The aims of this study were to develop a scale assessing nurses’ perceived skills and attitudes toward updated safety concepts, determine content validity, and examine internal consistency of the scale and subscales. Understanding nurses’ perceived skills and attitudes about safety concepts can be used in targeting strategies to enhance their safety practices. PMID:27479518

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

International Nuclear Information System (INIS)

Williams, R.A.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2011-11-01

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

Assessment of the long-term safety for SFR

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Selection of initial events of accelerator driven subcritical system

International Nuclear Information System (INIS)

Wang Qianglong; Hu Liqin; Wang Jiaqun; Li Yazhou; Yang Zhiyi

2013-01-01

The Probabilistic Safety Assessment (PSA) is an important tool in reactor safety analysis and a significant reference to the design and operation of reactor. It is the origin and foundation of the PSA for a reactor to select the initial events. Accelerator Driven Subcritical System (ADS) has advanced design characteristics, complicated subsystems and little engineering and operating experience, which makes it much more difficult to identify the initial events of ADS. Based on the current design project of ADS, the system's safety characteristics and special issues were analyzed in this article. After a series of deductions with Master Logic Diagram (MLD) and considering the relating experience of other advanced research reactors, a preliminary initial events was listed finally, which provided the foundation for the next safety assessment. (authors)

Animal-Free Chemical Safety Assessment

Directory of Open Access Journals (Sweden)

George D Loizou

2016-07-01

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

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

International Nuclear Information System (INIS)

2011-12-01

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

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

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

Safety Assessment for Decommissioning of Nuclear Facilities - From Methodology to the Use of Results in Decision Making

International Nuclear Information System (INIS)

Batandjieva, B.; Ferch, R.; Joubert, A.; Kaulard, J.; Manson, P.; Percival, K.; Thierfeldt, St.

2008-01-01

approach for reviewing safety assessments for decommissioning activities and as a basis for the regulatory decision making; - To provide a forum for exchange of experience in evaluation and demonstration of safety during decommissioning of various types of nuclear facilities. The project has involved over one hundred interested experts, of which fifty were actively involved throughout, from thirty plus countries working in eight working groups over the period 2004-2007. The experts presented and analyzed their experiences and on this basis developed recommendations on good practice in the development and review of safety assessments for decommissioning. According to the objectives, the project initially focused on the development of a harmonized methodology to undertake safety assessment. Later on the project entailed working groups testing out the methodology by applying it to a number of test cases representative of the typical facilities presently undergoing decommissioning. These included a nuclear power plant, a research reactor and a plutonium laboratory. The outcomes from the project and the key lessons learned are presented below. (authors)

Probabilistic safety assessment of the Fugen NPS

International Nuclear Information System (INIS)

Sotsu, Masutake; Iguchi, Yukihiro; Mizuno, Kouichi; Sato, Shinichirou; Shimizu, Miwako

1999-01-01

We performed a probabilistic safety assessment (PSA) on the Fugen NPS. The main topic of assessment was internal factors. We assessment core damage frequency (level 1 PSA) and containment damage frequency (level 2 PSA) during rated operation, and core damage frequency during shutdown (PSA during shutdowns). Our assessment showed that the core damage frequency of Fugen is well below the IAEA criteria for existing plants, that the conditional containment damage during shutdown is almost the target value of 0.1, and that the core damage frequency during shutdown is almost the same as that assessed during operation. These results confirm that the Fugen plant maintains a sufficient safety margin during shutdowns for regular inspections and for refueling. We developed and verified the effectiveness of an accident management plan incorporating the results of the assessment. (author)

Climate change and landscape development in post-closure safety assessment of solid radioactive waste disposal: Results of an initiative of the IAEA.

Science.gov (United States)

Lindborg, T; Thorne, M; Andersson, E; Becker, J; Brandefelt, J; Cabianca, T; Gunia, M; Ikonen, A T K; Johansson, E; Kangasniemi, V; Kautsky, U; Kirchner, G; Klos, R; Kowe, R; Kontula, A; Kupiainen, P; Lahdenperä, A-M; Lord, N S; Lunt, D J; Näslund, J-O; Nordén, M; Norris, S; Pérez-Sánchez, D; Proverbio, A; Riekki, K; Rübel, A; Sweeck, L; Walke, R; Xu, S; Smith, G; Pröhl, G

2018-03-01

The International Atomic Energy Agency has coordinated an international project addressing climate change and landscape development in post-closure safety assessments of solid radioactive waste disposal. The work has been supported by results of parallel on-going research that has been published in a variety of reports and peer reviewed journal articles. The project is due to be described in detail in a forthcoming IAEA report. Noting the multi-disciplinary nature of post-closure safety assessments, here, an overview of the work is given to provide researchers in the broader fields of radioecology and radiological safety assessment with a review of the work that has been undertaken. It is hoped that such dissemination will support and promote integrated understanding and coherent treatment of climate change and landscape development within an overall assessment process. The key activities undertaken in the project were: identification of the key processes that drive environmental change (mainly those associated with climate and climate change), and description of how a relevant future may develop on a global scale; development of a methodology for characterising environmental change that is valid on a global scale, showing how modelled global changes in climate can be downscaled to provide information that may be needed for characterising environmental change in site-specific assessments, and illustrating different aspects of the methodology in a number of case studies that show the evolution of site characteristics and the implications for the dose assessment models. Overall, the study has shown that quantitative climate and landscape modelling has now developed to the stage that it can be used to define an envelope of climate and landscape change scenarios at specific sites and under specific greenhouse-gas emissions assumptions that is suitable for use in quantitative post-closure performance assessments. These scenarios are not predictions of the future, but

Assessment of thermal fatigue crack propagation in safety injection PWR lines

International Nuclear Information System (INIS)

Simos, N.; Reich, M.; Costantino, C.J.; Hartzman, M.

1990-01-01

Cyclic thermal stratification resulting in alternating thermal stresses in pipe cross sections has been identified as the primary cause of high cycle thermal fatigue failure. A number of piping lines in operating plants around the world, susceptible to thermal stratification, have experienced circumferential cracking as a result of high levels of alternating bending stresses. This paper addresses the mechanisms of crack initiation and crack growth and provides estimates of fatigue cycles to failure for a typical safety injection line with such cyclic load history. Utilizing a 3-D finite element analysis, the temperature profile and the corresponding thermal stress field of a complete thermal cycle in a safety injection line consisting of a horizontal pipe section and an elbow, is obtained. Since the observed cracking occurred in the region of the elbow-to-horizontal pipe weld, the analysis performed assessed (1) the impact of the level of local geometric discontinuities on the initiation of an inside surface flaw is greatest and (2) the number of thermal cycles required to drive a small surface crack through the pipe wall. 12 refs., 14 figs., 2 tabs

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

International Nuclear Information System (INIS)

Metcalf, P.

2003-01-01

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

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

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

Fusion blanket inherent safety assessment

International Nuclear Information System (INIS)

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

1986-01-01

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

Marked point process framework for living probabilistic safety assessment and risk follow-up

International Nuclear Information System (INIS)

Arjas, Elja; Holmberg, Jan

1995-01-01

We construct a model for living probabilistic safety assessment (PSA) by applying the general framework of marked point processes. The framework provides a theoretically rigorous approach for considering risk follow-up of posterior hazards. In risk follow-up, the hazard of core damage is evaluated synthetically at time points in the past, by using some observed events as logged history and combining it with re-evaluated potential hazards. There are several alternatives for doing this, of which we consider three here, calling them initiating event approach, hazard rate approach, and safety system approach. In addition, for a comparison, we consider a core damage hazard arising in risk monitoring. Each of these four definitions draws attention to a particular aspect in risk assessment, and this is reflected in the behaviour of the consequent risk importance measures. Several alternative measures are again considered. The concepts and definitions are illustrated by a numerical example

Mathematical Safety Assessment Approaches for Thermal Power Plants

Directory of Open Access Journals (Sweden)

Zong-Xiao Yang

2014-01-01

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

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

A Computer Program for Assessing Nuclear Safety Culture Impact

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

OpenAIRE

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

2016-01-01

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

Safety assessment of the liquid-fed ceramic melter process

International Nuclear Information System (INIS)

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

1980-08-01

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

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

International Nuclear Information System (INIS)

Edwards, Gareth

2005-01-01

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

LANL Safety Conscious Work Environment (SCWE) Self-Assessment

Energy Technology Data Exchange (ETDEWEB)

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

2014-01-29

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

Quantitative risk assessment of digitalized safety systems

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Safety assessment of high consequence robotics system

International Nuclear Information System (INIS)

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

1996-01-01

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

Integrated Initiating Event Performance Indicators

International Nuclear Information System (INIS)

S. A. Eide; Dale M. Rasmuson; Corwin L. Atwood

2005-01-01

The U.S. Nuclear Regulatory Commission Industry Trends Program (ITP) collects and analyses industry-wide data, assesses the safety significance of results, and communicates results to Congress and other stakeholders. This paper outlines potential enhancements in the ITP to comprehensively cover the Initiating Events Cornerstone of Safety. Future work will address other cornerstones of safety. The proposed Tier 1 activity involves collecting data on ten categories of risk-significant initiating events, trending the results, and comparing early performance with prediction limits (allowable numbers of events, above which NRC action may occur). Tier 1 results would be used to monitor industry performance at the level of individual categories of initiating events. The proposed Tier 2 activity involves integrating the information for individual categories of initiating events into a single risk-based indicator, termed the Baseline Risk Index for Initiating Events or BRIIE. The BRIIE would be evaluated yearly and compared against a threshold. BRIIE results would be reported to Congress on a yearly basis

Farmworkers at the border: a bilingual initiative for occupational health and safety.

Science.gov (United States)

Acosta, Martha Soledad Vela; Sechrest, Lee; Chen, Mei-Kuang

2009-01-01

Bilingual and bicultural occupational health and safety interventions for Hispanic farmworkers are extremely rare and, because of language barriers and cultural differences, issues important to their health and safety on the job remain unaddressed. We designed, conducted, and assessed the first bilingual occupational health and safety education program for farmworkers attending High School Equivalency Programs (HEPs). We took an interdisciplinary participatory approach by integrating educators and researchers with a community advisory board to guide development, evaluation, and implementation of Work Safely-Trabaje con Cuidado Curriculum (Curriculum), a bilingual occupational health and safety curriculum. We created a quasi-experimental design using mixed-method evaluation (quantitative and qualitative elements) via pre- and posttest comparisons, follow-up surveys, and focus groups assessing the Curriculum effect on knowledge, safety risk perception (SRP), and safety behavior. Focus groups and follow-up surveys reflected success and acceptance of the Curriculum among participating farmworkers under the study's logic model. Completion of the Curriculum resulted in statistically significant improvements in the combined score of knowledge and SRP at the posttest (p = 0.001) and follow-up survey (p = 0.02) in the intervention group. After completing this study, the Curriculum was permanently adopted by the two high school equivalency sites involved. The participatory approach resulted in integration of community and applied research partnership. The potential to expand use of this Curriculum by other HEP sites can further assess effectiveness and external validity among underserved minority groups.

Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

2009-01-01

After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

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

International Nuclear Information System (INIS)

Xiang Yuanyuan; Chen Xukun; Xu Rongbin

2012-01-01

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

Safety/security interface assessments at commercial nuclear power plants

International Nuclear Information System (INIS)

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

1985-01-01

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

Safety/security interface assessments at commercial nuclear power plants

International Nuclear Information System (INIS)

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

1985-07-01

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

Examining the influence of urban definition when assessing relative safety of drinking-water in Nigeria.

Science.gov (United States)

Christenson, Elizabeth; Bain, Robert; Wright, Jim; Aondoakaa, Stephen; Hossain, Rifat; Bartram, Jamie

2014-08-15

Reducing inequalities is a priority from a human rights perspective and in water and public health initiatives. There are periodic calls for differential national and global standards for rural and urban areas, often justified by the suggestion that, for a given water source type, safety is worse in urban areas. For instance, initially proposed post-2015 water targets included classifying urban but not rural protected dug wells as unimproved. The objectives of this study were to: (i) examine the influence of urban extent definition on water safety in Nigeria, (ii) compare the frequency of thermotolerant coliform (TTC) contamination and prevalence of sanitary risks between rural and urban water sources of a given type and (iii) investigate differences in exposure to contaminated drinking-water in rural and urban areas. We use spatially referenced data from a Nigerian national randomized sample survey of five improved water source types to assess the extent of any disparities in urban-rural safety. We combined the survey data on TTC and sanitary risk with map layers depicting urban versus rural areas according to eight urban definitions. When examining water safety separately for each improved source type, we found no significant urban-rural differences in TTC contamination and sanitary risk for groundwater sources (boreholes and protected dug wells) and inconclusive findings for piped water and stored water. However, when improved and unimproved source types were combined, TTC contamination was 1.6 to 2.3 times more likely in rural compared to urban water sources depending on the urban definition. Our results suggest that different targets for urban and rural water safety are not justified and that rural dwellers are more exposed to unsafe water than urban dwellers. Additionally, urban-rural analyses should assess multiple definitions or indicators of urban to assess robustness of findings and to characterize a gradient that disaggregates the urban-rural dichotomy

Safety assessment of biotechnology-derived pharmaceuticals: ICH and beyond.

Science.gov (United States)

Serabian, M A; Pilaro, A M

1999-01-01

Many scientific discussions, especially in the past 8 yr, have focused on definition of criteria for the optimal assessment of the preclinical toxicity of pharmaceuticals. With the current overlap of responsibility among centers within the Food and Drug Administration (FDA), uniformity of testing standards, when appropriate, would be desirable. These discussions have extended beyond the boundaries of the FDA and have culminated in the acceptance of formalized, internationally recognized guidances. The work of the International Committee on Harmonisation (ICH) and the initiatives developed by the FDA are important because they (a) represent a consensus scientific opinion, (b) promote consistency, (c) improve the quality of the studies performed, (d) assist the public sector in determining what may be generally acceptable to prepare product development plans, and (e) provide guidance for the sponsors in the design of preclinical toxicity studies. Disadvantages associated with such initiatives include (a) the establishment of a historical database that is difficult to relinquish, (b) the promotion of a check-the-box approach, i.e., a tendancy to perform only the minimum evaluation required by the guidelines, (c) the creation of a disincentive for industry to develop and validate new models, and (d) the creation of state-of-the-art guidances that may not allow for appropriate evaluation of novel therapies. The introduction of biotechnology-derived pharmaceuticals for clinical use has often required the application of unique approaches to assessing their safety in preclinical studies. There is much diversity among these products, which include the gene and cellular therapies, monoclonal antibodies, human-derived recombinant regulatory proteins, blood products, and vaccines. For many of the biological therapies, there will be unique product issues that may require specific modifications to protocol design and may raise additional safety concerns (e.g., immunogenicity

Safety assessment for Generation IV nuclear systems

International Nuclear Information System (INIS)

Leahy, T.J.

2012-01-01

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

Nirex safety assessment research programme: 1987/88

International Nuclear Information System (INIS)

George, D.; Hodgkinson, D.P.

1987-01-01

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

Safety assessment requirements for onsite transfers of radioactive material

International Nuclear Information System (INIS)

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

1992-05-01

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

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/

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

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

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)

Modeling issues associated with production reactor safety assessment

International Nuclear Information System (INIS)

Stack, D.W.; Thomas, W.R.

1990-01-01

This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

Assessing safety culture using RADAR matrix

International Nuclear Information System (INIS)

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

2009-01-01

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

Developing IAM for Life Cycle Safety Assessment

NARCIS (Netherlands)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-01

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

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

NARCIS (Netherlands)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

1998-01-01

The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

Fire safety assessment of tunnel structures

DEFF Research Database (Denmark)

Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

2011-01-01

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

Safety management system needs assessment.

Science.gov (United States)

2016-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

Safety and security risk assessments--now demystified!

Science.gov (United States)

White, Donald E

2011-01-01

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

Method to Find Recovery Event Combinations in Probabilistic Safety Assessment

International Nuclear Information System (INIS)

Jung, Woo Sik; Riley, Jeff

2016-01-01

These research activities may develop mathematical methods, engineering analyses, and business processes. The research activities of the project covered by this scope are directed toward the specific issues of implementing the methods and strategies on a computational platform, identifying the features and enhancements to EPRI tools that would be necessary to realize significant improvements to the risk assessments performed by the end user. Fault tree analysis is extensively and successfully applied to the risk assessment of safety-critical systems such as nuclear, chemical and aerospace systems. The fault tree analysis is being used together with an event tree analysis in PSA of nuclear power plants. Fault tree solvers for a PSA are mostly based on the cutset-based algorithm. They generate minimal cut sets (MCSs) from a fault tree. The most popular fault tree solver in the PSA industry is FTREX. During the course of this project, certain technical issues (see Sections 2 to 5) have been identified that need to be addressed regarding how minimal cut sets are generated and quantified. The objective of this scope of the work was to develop new methods or techniques to address these technical limitations. By turning on all the cutset initiators (%1, %2, %3, %), all the possible minimal cut sets can be calculated easier than with the original fault tree. It is accomplished by the fact that the number of events in the minimal cut sets are significantly reduced by using cutset initiators instead of random failure events. And byy turning on a few chosen cutset initiators and turning off the other cutset initiators, minimal cut sets of the selected cutset initiator(s) can be easily calculated. As explained in the previous Sections, there is no way to calculate these minimal cut sets by turning off/on the random failure events in the original fault tree

Method to Find Recovery Event Combinations in Probabilistic Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Jung, Woo Sik [Sejong University, Seoul (Korea, Republic of); Riley, Jeff [Electric Power Research, Palo Alto (United States)

2016-05-15

These research activities may develop mathematical methods, engineering analyses, and business processes. The research activities of the project covered by this scope are directed toward the specific issues of implementing the methods and strategies on a computational platform, identifying the features and enhancements to EPRI tools that would be necessary to realize significant improvements to the risk assessments performed by the end user. Fault tree analysis is extensively and successfully applied to the risk assessment of safety-critical systems such as nuclear, chemical and aerospace systems. The fault tree analysis is being used together with an event tree analysis in PSA of nuclear power plants. Fault tree solvers for a PSA are mostly based on the cutset-based algorithm. They generate minimal cut sets (MCSs) from a fault tree. The most popular fault tree solver in the PSA industry is FTREX. During the course of this project, certain technical issues (see Sections 2 to 5) have been identified that need to be addressed regarding how minimal cut sets are generated and quantified. The objective of this scope of the work was to develop new methods or techniques to address these technical limitations. By turning on all the cutset initiators (%1, %2, %3, %), all the possible minimal cut sets can be calculated easier than with the original fault tree. It is accomplished by the fact that the number of events in the minimal cut sets are significantly reduced by using cutset initiators instead of random failure events. And byy turning on a few chosen cutset initiators and turning off the other cutset initiators, minimal cut sets of the selected cutset initiator(s) can be easily calculated. As explained in the previous Sections, there is no way to calculate these minimal cut sets by turning off/on the random failure events in the original fault tree.

Visualization of Safety Assessment Result Using GIS in SITES

International Nuclear Information System (INIS)

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

2006-01-01

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

Safety Margin Assessment (SM2A): Stimulation for Further Development of BEPU Approaches?

International Nuclear Information System (INIS)

Zimmermann, Martin A.

2013-01-01

During recent years, many nuclear power plants underwent significant modifications, e.g. power up-rating. While compliance with all the deterministic acceptance criteria must be shown during the licensing process, the larger core inventory and the facts that the plant response might get closer to the limits after a power up-rate, suggest an increase of the core damage frequency (CDF) and other possible risk indicators. Hence, a framework to quantitatively assess a change in plant safety margin becomes very desirable. The Committee on the Safety of Nuclear Installations (CSNI) mandated the Safety Margin Action Plan expert group (SMAP) to develop a framework for the assessment of such changes to safety margin. This framework combines PSA and the analytical techniques developed in BEPU. CSNI then mandated the SM2A expert group to especially explore the practicability of the SMAP framework. This pilot study was completed end of 2010. An increase of the (conditional) probability of exceedance for a surrogate acceptance limit (PCT) indicating core damage was successfully evaluated for the selected sequences from several initiating event trees, and it was found that only a restricted number of sequences need to be analyzed. Based on the insights gained from this study, areas of methodology improvement have been identified and related proposals for further R and D work will be discussed. (authors)

Assessment of Patient Safety Culture in Primary Health Care Settings in Kuwait

Directory of Open Access Journals (Sweden)

Maha Mohamed Ghobashi

2014-01-01

Full Text Available Background Patient safety is critical component of health care quality. We aimed to assess the awareness of primary healthcare staff members about patient safety culture and explore the areas of deficiency and opportunities for improvement concerning this issue.Methods: This descriptive cross sectional study surveyed 369 staff members in four primary healthcare centers in Kuwait using self-administered “Hospital Survey on Patient Safety Culture” adopted questionnaire. The total number of respondents was 276 participants (response rate = 74.79%.Results: Five safety dimensions with lowest positivity (less than 50% were identified and these are; the non – punitive response to errors, frequency of event reporting, staffing, communication openness, center handoffs and transitions with the following percentages of positivity 24%, 32%, 41%, 45% and 47% respectively. The dimensions of highest positivity were teamwork within the center’s units (82% and organizational learning (75%.Conclusion: Patient safety culture in primary healthcare settings in Kuwait is not as strong as improvements for the provision of safe health care. Well-designed patient safety initiatives are needed to be integrated with organizational policies, particularly the pressing need to address the bioethical component of medical errors and their disclosure, communication openness and emotional issues related to them and investing the bright areas of skillful organizational learning and strong team working attitudes.    

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

Initial Assessment of Whiplash Patients

Directory of Open Access Journals (Sweden)

R Gunzburg

2003-01-01

Full Text Available The article looks at how for severe trauma, the outcome of treatment depends on the initial medical care. This has now also been accepted for whiplash associated disorders, underlining the importance of a proper initial assessment. Once major injury has been excluded and the diagnosis of whiplash associated disorder has been established, the initial treatment of whiplash in the emergency room can be started. The four key points to remember are described, including reassuring the patient about evolution, no soft collar, nonsteroidal anti-inflammatory drugs and early mobilisation.

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

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

Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

Energy Technology Data Exchange (ETDEWEB)

Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

Assessment of safety regulation using an artificial society

International Nuclear Information System (INIS)

Furuta, Kazuo; Nagase, Masaya

2005-01-01

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

Continuous Assessment of Safety Margin for the 14-MW TRIGA Reactor

International Nuclear Information System (INIS)

Ciocanescu, M.; Georgescu, D.; Doru, O.

2008-01-01

The assessment of reactor safety implies analyses of the reactor and its systems response to a range of postulated initiating events (such as malfunction or failures of equipment, operator errors, external events and so on which could lead to either anticipated operational occurrences or to accident conditions. Decreasing in heat removal by the reactor cooling system may be considered as a process disturbance which may lead to a postulated initiating event. The cold source for the reactor cooling system, in case of TRIGA-14 MW reactor is the atmosphere by the secondary cooling towers. The ability to evacuate the heat produced by the reactor core ranges between the outlet temperature of the core flow and the outdoors temperature in air, which is subject to season and day variation. Selected values for safety limits, safety system settings and limiting condition(s) are derived from safety analysis and are consistent with the operational state of the reactor. When a limiting condition for safe operation is not satisfied, the operating personal is supposed to take the appropriate action(s) to ensure safety. Operating requirements and the safety system are presented. The reactor operating safety parameters from the main Data Acquisition System are transferred to an AT personal computer. These selected parameters are the following: - average inlet temperature which is calculated as an average temperature measured by 20 type K thermocouples distributed within a 4 x 5 matrix located on the top of the reactor core; - average outlet temperature which is calculated as an average record from 10 type K thermocouples placed in the outlet pipe; - average flow rate which is calculated as an average value from four transducers (two for the inlet flow rate and two for the outlet flow rate). Due to its high instability, this value is also filtered using a two-pole low-pass filter (software); - reactor thermal power value derivable from the previous parameters or obtained from the

Integrated Deterministic-Probabilistic Safety Assessment Methodologies

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-01

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

Human reliability in probabilistic safety assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

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

Probabilistic assessment of NPP safety under aircraft impact

International Nuclear Information System (INIS)

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

1999-01-01

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

Nirex Safety Assessment Research Programme bibliography, 1990

International Nuclear Information System (INIS)

Cooper, M.J.

1990-10-01

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

Understanding and assessing safety culture

International Nuclear Information System (INIS)

Dalling, Ian

1997-01-01

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

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

Directory of Open Access Journals (Sweden)

Ernest S. Fung

2018-06-01

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

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

International Nuclear Information System (INIS)

Birkhofer, A.

1992-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Savannah River Site K-Reactor Probabilistic Safety Assessment

International Nuclear Information System (INIS)

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

1992-12-01

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

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

Science.gov (United States)

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

2015-06-01

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

Suggestions on the Development of Safety Culture Assessment Method

International Nuclear Information System (INIS)

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

2006-01-01

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

Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

2014-01-01

As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia

Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mazleha Maskin; Phongsakorn, P.T.; Tonny, A.L.; Fedrick, C.M.B.; Faizal Mohamed; Mohamad Fauzi Saad; Ahmad Razali Ismail; Mohamad Puad Haji Abu

2013-01-01

Full-text: As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia. (author)

Implementation of a risk assessment tool based on a probabilistic safety assessment developed for radiotherapy practices

International Nuclear Information System (INIS)

Paz, A.; Godinez, V.; Lopez, R.

2010-10-01

The present work describes the implementation process and main results of the risk assessment to the radiotherapy practices with Linear Accelerators (Linac), with cobalt 60, and with brachytherapy. These evaluations were made throughout the risk assessment tool for radiotherapy practices SEVRRA (risk evaluation system for radiotherapy), developed at the Mexican National Commission in Nuclear Safety and Safeguards derived from the outcome obtained with the Probabilistic Safety Analysis developed at the Ibero-American Regulators Forum for these radiotherapy facilities. The methodology used is supported by risk matrices method, a mathematical tool that estimates the risk to the patient, radiation workers and public from mechanical failures, mis calibration of the devices, human mistakes, and so. The initiating events are defined as those undesirable events that, together with other failures, can produce a delivery of an over-dose or an under-dose of the medical prescribed dose, to the planned target volume, or a significant dose to non prescribed human organs. Initiating events frequency and reducer of its frequency (actions intended to avoid the accident) are estimated as well as robustness of barriers to those actions, such as mechanical switches, which detect and prevent the accident from occurring. The spectrum of the consequences is parameterized, and the actions performed to reduce the consequences are identified. Based on this analysis, a software tool was developed in order to simplify the evaluations to radiotherapy installations and it has been applied as a first step forward to some Mexican installations, as part of a national implementation process, the final goal is evaluation of all Mexican facilities in the near future. The main target and benefits of the SEVRRA implementation are presented in this paper. (Author)

Implementation of a risk assessment tool based on a probabilistic safety assessment developed for radiotherapy practices

Energy Technology Data Exchange (ETDEWEB)

Paz, A.; Godinez, V.; Lopez, R., E-mail: abpaz@cnsns.gob.m [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2010-10-15

The present work describes the implementation process and main results of the risk assessment to the radiotherapy practices with Linear Accelerators (Linac), with cobalt 60, and with brachytherapy. These evaluations were made throughout the risk assessment tool for radiotherapy practices SEVRRA (risk evaluation system for radiotherapy), developed at the Mexican National Commission in Nuclear Safety and Safeguards derived from the outcome obtained with the Probabilistic Safety Analysis developed at the Ibero-American Regulators Forum for these radiotherapy facilities. The methodology used is supported by risk matrices method, a mathematical tool that estimates the risk to the patient, radiation workers and public from mechanical failures, mis calibration of the devices, human mistakes, and so. The initiating events are defined as those undesirable events that, together with other failures, can produce a delivery of an over-dose or an under-dose of the medical prescribed dose, to the planned target volume, or a significant dose to non prescribed human organs. Initiating events frequency and reducer of its frequency (actions intended to avoid the accident) are estimated as well as robustness of barriers to those actions, such as mechanical switches, which detect and prevent the accident from occurring. The spectrum of the consequences is parameterized, and the actions performed to reduce the consequences are identified. Based on this analysis, a software tool was developed in order to simplify the evaluations to radiotherapy installations and it has been applied as a first step forward to some Mexican installations, as part of a national implementation process, the final goal is evaluation of all Mexican facilities in the near future. The main target and benefits of the SEVRRA implementation are presented in this paper. (Author)

Safety assessment of primary system components at the USNRC

Energy Technology Data Exchange (ETDEWEB)

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

1988-12-31

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

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

International Nuclear Information System (INIS)

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

2008-12-01

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

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

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

Flamanville 3 EPR, Safety Assessment and On-site Inspections

International Nuclear Information System (INIS)

Piedagnel, Corinne; Tarallo, Francois; Monnot, Bernard

2011-01-01

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

IRIS guidelines. 2014 ed. Integrated Review of Infrastructure for Safety (IRIS) for self-assessment when establishing the safety infrastructure for a nuclear power programme

International Nuclear Information System (INIS)

2014-01-01

The IAEA safety standards reflect an international consensus on what constitutes a high level of safety for protecting people and the environment, and therefore represent what all Member States should achieve, whilst recognizing the ultimate responsibility of each State to ensure safety when implementing a nuclear power programme. IAEA Safety Standards Series No. SSG-16, entitled Establishing the Safety Infrastructure for a Nuclear Power Programme was published in order to provide recommendations, presented in the form of sequential actions, on meeting safety requirements progressively during the initial three phases of the development of safety, as described in INSAG-22, Nuclear Safety Infrastructure for a National Nuclear Power Programme Supported by the IAEA Fundamental Safety Principles. To that end, the 200 safety related actions, which are proposed by SSG-16, constitute a roadmap to establish a foundation for promoting a high level of safety over the entire lifetime of the nuclear power plant. These actions reflect international consensus on good practice in order to achieve full implementation of IAEA safety standards. The IAEA has developed a methodology and tool, the Integrated Review of Infrastructure for Safety (IRIS), to assist States in undertaking self-assessment with respect to SSG-16 recommendations when establishing the safety infrastructure for a nuclear power programme, and to develop an action plan for improvement. The IRIS methodology and the associated tool are fully compatible with the IAEA safety standards and are also used, when appropriate, in the preparation of review missions, such as the Integrated Regulatory Review Service and advisory missions. The present guidelines describe the IRIS methodology for self-assessment against SSG-16 recommendations. Through IRIS implementation, every organization concerned with nuclear safety may gain proper awareness and engage in a continuous progressive process to develop the effective national

Prospects for probabilistic safety assessment

International Nuclear Information System (INIS)

Hirschberg, S.

1992-01-01

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

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

International Nuclear Information System (INIS)

1984-05-01

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

Training courses on integrated safety assessment modelling for waste repositories

International Nuclear Information System (INIS)

Mallants, D.

2007-01-01

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

Human Reliability in Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

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

Interim process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Sellin, Patrick

2004-08-01

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

Interim process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrick (ed.)

2004-08-01

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

Probabilistic Safety Assessment 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.

Short-term initiatives for enhancing cyber-safety within South African schools

Directory of Open Access Journals (Sweden)

Elmarie Kritzinger

2016-07-01

Full Text Available The rate of technological development across the globe is dramatic. The decreasing cost and increasing availability of ICT devices means that its users are no longer exclusively industry or government employees – they are now also home users. Home users integrate ICT in their daily lives for education, socialising and information gathering. However, using ICT is associated with risks and threats, such as identity theft and phishing scams. Most home users of ICT do not have the necessary information technology and Internet skills to protect themselves and their information. School learners, in particular, are not sufficiently educated on how to use technological devices safely, especially in developing countries such as South Africa. The national school curriculum in South Africa currently does not make provision for cyber-safety education, and the availability of supporting material and training for ICT teachers in South Africa is limited, resulting in a lack of knowledge and skills regarding cyber-safety. The research in hand focuses on the situation concerning cyber-safety awareness in schools and has adopted a short-term approach towards cyber-safety among teachers and school learners in South Africa until a formal long-term national approach has been implemented. This study takes a quantitative approach to investigating the current options of teachers to enhance cyber-safety among learners in their schools. The research proposes that short-term initiatives (i.e. posters can increase learners’ awareness of cyber-safety until formal cyber-safety awareness methods have been introduced.

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Concepts and possibilities of fracture mechanics for fracture safety assessment

International Nuclear Information System (INIS)

Blauel, J.

1980-01-01

In very tough materials for pressure vessels and pipelines of nuclear plants, cracking begins in a stable manner and only after macroscopic plastic deformations and crack blunting. It is possible to describe this elasto-plastic fracture behaviour and to quantify the safety margin compared to the assessment criteria based on linear elastic stressing and initiation by the concept of the J integral, the crack peak width and the crack resistance Jsub(R) curve. The numerous problems of details still open and the partly very limited validity range should not prevent the further investigation into the great possibilities of this concept and making greater use of the interpretation of large scale tests. (orig./RW) [de

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)

Integrated Safety Assessment (ISA): An approach for the assessment of the software aspects of protection systems

International Nuclear Information System (INIS)

Izquierdo-Rocha, Jose Maria; Sanchez-Perea, Miguel; Cojazzi, Giacomo

2004-01-01

This paper reviews the main features of ISA, a concept developed as a result of previous work on safety assessment and dynamic reliability. The method links the dynamics of the facility with its operating environment, subject to transitions between different time evolutions due to failures and/or system/operator interventions. For situations dominated by Deterministic Transitions (i.e., transitions upon deterministic demands as a result for instance of exceeding automatic-actions/alarm setpoints), the methodology can be considered an extension of PSA and accident analysis techniques that replaces the static event tree with a Deterministic Dynamic Event Tree (DDET) concept based on the Theory of Probabilistic Dynamics. The paper also summarizes some results of an ISA application to the assessment of the Emergency Operating Procedure (EOP) of a PWR-W to mitigate the Steam Generator Tube Rupture (SGTR) initiating event. (author)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Report to the Attorney General on Body Armor Safety Initiative Testing and Activities

National Research Council Canada - National Science Library

2005-01-01

On November 17, 2003, Attorney General John Ashcroft announced the U.S. Department of Justice's Body Armor Safety Initiative in response to concerns from the law enforcement community regarding the effectiveness of body armor in use...

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

OpenAIRE

Nganje, William E.

1999-01-01

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

Safety re-assessment of AECL test and research reactors

International Nuclear Information System (INIS)

Winfield, D.J.

1990-01-01

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

Examining the influence of urban definition when assessing relative safety of drinking-water in Nigeria

Energy Technology Data Exchange (ETDEWEB)

Christenson, Elizabeth; Bain, Robert [The Water Institute at UNC, University of North Carolina at Chapel Hill, NC (United States); Wright, Jim [Geography and Environment, University of Southampton, Southampton (United Kingdom); Aondoakaa, Stephen [Geography and Environmental Management, University of Abuja, Abuja (Nigeria); School of Geography, University of Nottingham, Nottingham (United Kingdom); Hossain, Rifat [World Health Organization, Geneva (Switzerland); Bartram, Jamie, E-mail: jbartram@unc.edu [The Water Institute at UNC, University of North Carolina at Chapel Hill, NC (United States)

2014-08-15

Reducing inequalities is a priority from a human rights perspective and in water and public health initiatives. There are periodic calls for differential national and global standards for rural and urban areas, often justified by the suggestion that, for a given water source type, safety is worse in urban areas. For instance, initially proposed post-2015 water targets included classifying urban but not rural protected dug wells as unimproved. The objectives of this study were to: (i) examine the influence of urban extent definition on water safety in Nigeria, (ii) compare the frequency of thermotolerant coliform (TTC) contamination and prevalence of sanitary risks between rural and urban water sources of a given type and (iii) investigate differences in exposure to contaminated drinking-water in rural and urban areas. We use spatially referenced data from a Nigerian national randomized sample survey of five improved water source types to assess the extent of any disparities in urban–rural safety. We combined the survey data on TTC and sanitary risk with map layers depicting urban versus rural areas according to eight urban definitions. When examining water safety separately for each improved source type, we found no significant urban–rural differences in TTC contamination and sanitary risk for groundwater sources (boreholes and protected dug wells) and inconclusive findings for piped water and stored water. However, when improved and unimproved source types were combined, TTC contamination was 1.6 to 2.3 times more likely in rural compared to urban water sources depending on the urban definition. Our results suggest that different targets for urban and rural water safety are not justified and that rural dwellers are more exposed to unsafe water than urban dwellers. Additionally, urban–rural analyses should assess multiple definitions or indicators of urban to assess robustness of findings and to characterize a gradient that disaggregates the urban�

Examining the influence of urban definition when assessing relative safety of drinking-water in Nigeria

International Nuclear Information System (INIS)

Christenson, Elizabeth; Bain, Robert; Wright, Jim; Aondoakaa, Stephen; Hossain, Rifat; Bartram, Jamie

2014-01-01

Reducing inequalities is a priority from a human rights perspective and in water and public health initiatives. There are periodic calls for differential national and global standards for rural and urban areas, often justified by the suggestion that, for a given water source type, safety is worse in urban areas. For instance, initially proposed post-2015 water targets included classifying urban but not rural protected dug wells as unimproved. The objectives of this study were to: (i) examine the influence of urban extent definition on water safety in Nigeria, (ii) compare the frequency of thermotolerant coliform (TTC) contamination and prevalence of sanitary risks between rural and urban water sources of a given type and (iii) investigate differences in exposure to contaminated drinking-water in rural and urban areas. We use spatially referenced data from a Nigerian national randomized sample survey of five improved water source types to assess the extent of any disparities in urban–rural safety. We combined the survey data on TTC and sanitary risk with map layers depicting urban versus rural areas according to eight urban definitions. When examining water safety separately for each improved source type, we found no significant urban–rural differences in TTC contamination and sanitary risk for groundwater sources (boreholes and protected dug wells) and inconclusive findings for piped water and stored water. However, when improved and unimproved source types were combined, TTC contamination was 1.6 to 2.3 times more likely in rural compared to urban water sources depending on the urban definition. Our results suggest that different targets for urban and rural water safety are not justified and that rural dwellers are more exposed to unsafe water than urban dwellers. Additionally, urban–rural analyses should assess multiple definitions or indicators of urban to assess robustness of findings and to characterize a gradient that disaggregates the urban�

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

Use of the Home Safety Self-Assessment Tool (HSSAT) within Community Health Education to Improve Home Safety.

Science.gov (United States)

Horowitz, Beverly P; Almonte, Tiffany; Vasil, Andrea

2016-10-01

This exploratory research examined the benefits of a health education program utilizing the Home Safety Self-Assessment Tool (HSSAT) to increase perceived knowledge of home safety, recognition of unsafe activities, ability to safely perform activities, and develop home safety plans of 47 older adults. Focus groups in two senior centers explored social workers' perspectives on use of the HSSAT in community practice. Results for the health education program found significant differences between reported knowledge of home safety (p = .02), ability to recognize unsafe activities (p = .01), safely perform activities (p = .04), and develop a safety plan (p = .002). Social workers identified home safety as a major concern and the HSSAT a promising assessment tool. Research has implications for reducing environmental fall risks.

Probabilistic safety assessments of nuclear power plants for low power and shutdown modes

International Nuclear Information System (INIS)

2000-03-01

Within the past several years the results of nuclear power plant operating experience and performance of probabilistic safety assessments (PSAs) for low power and shutdown operating modes have revealed that the risk from operating modes other than full power may contribute significantly to the overall risk from plant operations. These early results have led to an increased focus on safety during low power and shutdown operating modes and to an increased interest of many plant operators in performing shutdown and low power PSAs. This publication was developed to provide guidance and insights on the performance of PSA for shutdown and low power operating modes. The preparation of this publication was initiated in 1994. Two technical consultants meetings were conducted in 1994 and one in February 1999 in support of the development of this report

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

The current status of exposure-driven approaches for chemical safety assessment: A cross-sector perspective.

Science.gov (United States)

Sewell, Fiona; Aggarwal, Manoj; Bachler, Gerald; Broadmeadow, Alan; Gellatly, Nichola; Moore, Emma; Robinson, Sally; Rooseboom, Martijn; Stevens, Alexander; Terry, Claire; Burden, Natalie

2017-08-15

For the purposes of chemical safety assessment, the value of using non-animal (in silico and in vitro) approaches and generating mechanistic information on toxic effects is being increasingly recognised. For sectors where in vivo toxicity tests continue to be a regulatory requirement, there has been a parallel focus on how to refine studies (i.e. reduce suffering and improve animal welfare) and increase the value that in vivo data adds to the safety assessment process, as well as where to reduce animal numbers where possible. A key element necessary to ensure the transition towards successfully utilising both non-animal and refined safety testing is the better understanding of chemical exposure. This includes approaches such as measuring chemical concentrations within cell-based assays and during in vivo studies, understanding how predicted human exposures relate to levels tested, and using existing information on human exposures to aid in toxicity study design. Such approaches promise to increase the human relevance of safety assessment, and shift the focus from hazard-driven to risk-driven strategies similar to those used in the pharmaceutical sectors. Human exposure-based safety assessment offers scientific and 3Rs benefits across all sectors marketing chemical or medicinal products. The UK's National Centre for the Replacement, Refinement and Reduction of Animals in Research (NC3Rs) convened an expert working group of scientists across the agrochemical, industrial chemical and pharmaceutical industries plus a contract research organisation (CRO) to discuss the current status of the utilisation of exposure-driven approaches, and the challenges and potential next steps for wider uptake and acceptance. This paper summarises these discussions, highlights the challenges - particularly those identified by industry - and proposes initial steps for moving the field forward. Copyright © 2017 The Author(s). Published by Elsevier B.V. All rights reserved.

Assessing progress in the development of safety culture

International Nuclear Information System (INIS)

Rotaru, Ioan; Ghita, Sorin

1999-01-01

The concept of safety culture was introduced by the International Nuclear Safety Advisory Group (INSAG) in the Summary Report on the Post-Accident Meeting on the Chernobyl Accident in 1986. The concept was further expanded in the 1988 INSAG-3 report, Basic Safety Principles for Nuclear Power Plants, and again in 1991 in the INSAG-4 report. Recognizing the increasing role that safety culture is expected to play in nuclear installations worldwide, the Convention on Nuclear Safety states the Contracting Parties' desire 'to promote an effective nuclear safety culture'. The concept of safety culture is defined in INSAG-4 as follows: Safety culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. Safety culture is also an amalgamation of values, standards, morals and norms of acceptable behaviour. These are aimed at maintaining a self disciplined approach to the enhancement of safety beyond legislative and regulatory requirements. Therefore, the safety culture has to be inherent in the thoughts and actions of all the individuals at every level in an organization. The leadership provided by top management is crucial. Safety culture applies to conventional and personal safety as well as nuclear safety. All safety consideration are affected by common points of beliefs, attitudes, behaviour, and cultural differences, closely linked to a shared system of values and standards. The paper poses questions and tries to find answers relative to issues like: - how to assess progress; - specific organizational indicators of a progressive safety culture; - detection of incipient weaknesses in safety culture (organizational issues, employee issues, technology issues); - revitalizing a weakened safety culture; - overall assesment of safety culture; - general evaluation model. In conclusion, there is no consistent and

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-15

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

Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities

International Nuclear Information System (INIS)

Batandjieva, B.; Torres-Vidal, C.

2002-01-01

The International Atomic Energy Agency (IAEA) Coordinated research program ''Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities'' (ISAM) has developed improved safety assessment methodology for near surface disposal facilities. The program has been underway for three years and has included around 75 active participants from 40 countries. It has also provided examples for application to three safety cases--vault, Radon type and borehole radioactive waste disposal facilities. The program has served as an excellent forum for exchange of information and good practices on safety assessment approaches and methodologies used worldwide. It also provided an opportunity for reaching broad consensus on the safety assessment methodologies to be applied to near surface low and intermediate level waste repositories. The methodology has found widespread acceptance and the need for its application on real waste disposal facilities has been clearly identified. The ISAM was finalized by the end of 2000, working material documents are available and an IAEA report will be published in 2002 summarizing the work performed during the three years of the program. The outcome of the ISAM program provides a sound basis for moving forward to a new IAEA program, which will focus on practical application of the safety assessment methodologies to different purposes, such as licensing radioactive waste repositories, development of design concepts, upgrading existing facilities, reassessment of operating repositories, etc. The new program will also provide an opportunity for development of guidance on application of the methodology that will be of assistance to both safety assessors and regulators

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

Science.gov (United States)

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

2017-11-01

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

[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

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

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

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of