Preliminary safety evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)

International Nuclear Information System (INIS)

Dunn, T.D.; Lommers, L.J.; Tangirala, V.E.

1994-04-01

A qualitative comparison between the safety characteristics of the Gas Turbine-Modular Helium Reactor (GT-MHR) and those of the steam cycle shows that the two designs achieve equivalent levels of overall safety performance. This comparison is obtained by applying the scaling laws to detailed steam-cycle computations as well as the conclusions obtained from preliminary GT-MHR model simulations. The gas turbine design is predicted to be superior for some event categories, while the steam cycle design is better for others. From a safety perspective, the GT-MHR has a modest advantage for pressurized conduction cooldown events. Recent computational simulations of 102 column, 550 MW(t) GT-MHR during a depressurized conduction cooldown show that peak fuel temperatures are within the limits. The GT-MHR has a significantly lower risk due to water ingress events under operating conditions. Two additional scenarios, namely loss of load event and turbine deblading event that are specific to the GT-MHR design are discussed. Preliminary evaluation of the GT-MHR's safety characteristics indicate that the GT-MHR can be expected to satisfy or exceed its safety requirements

Preliminary Study on the Revision of Nuclear Safety Policy Statement

International Nuclear Information System (INIS)

Lee, Y. E.; Lee, S. H.; Chang, H. S.; Choi, K. S.; Jung, S. J.

2011-01-01

Nuclear safety policy in Korea is currently declared in the Nuclear Safety Charter as the highest tier document and safety principles and directions are announced in the Nuclear Safety Policy Statement. As the circumstances affecting on the nuclear safety policy change, it needs to revise the Statement. This study aims to develop the revised Nuclear Safety Policy Statement to declare that securing safety is a prerequisite to the utilization of nuclear energy, and that all workers in nuclear industry and regulatory body must adhere to the principle of priority to safety. As a result, two different types of revision are being prepared as of August. One is based on the spirit of Nuclear Safety Charter as well as the direction of future-oriented safety policies including the changes in the environment after declaration of the Statement. The other is to declare the fundamental safety objective and safety principles as the top philosophy of national nuclear safety policy by adopting the '10 Safety Principles in IAEA Safety Fundamental' instead of the current Charter. Both versions of revision are subject to further in-depth discussion. However once the revision is finalized and declared, it would be useful to accomplish effectively the organizational responsibilities and to enhance the public confidence in nuclear safety by performing the regulatory activities in a planned and systematic manner and promulgating the government's dedication to priority to safety

Preliminary safety evaluation for the spent nuclear fuel project`s cold vacuum drying system

Energy Technology Data Exchange (ETDEWEB)

Garvin, L.J., Westinghouse Hanford

1996-07-01

This preliminary safety evaluation (PSE) considers only the Cold Vacuum Drying System (CVDS) facility and its mission as it relates to the integrated process strategy (WHC 1995). The purpose of the PSE is to identify those CBDS design functions that may require safety- class and safety-significant accident prevention and mitigation features.

Preliminary Study on the Development of Quantitative Safety Culture Index

International Nuclear Information System (INIS)

Lee, Young Eal; Kim, Hun Sil; Ahn, Nam Sung

2005-01-01

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. Because it needs to be recognized as the most significant consciousness to achieve the nuclear safety performance, Korean government and nuclear power generation company have tried to develop the practical method to improve the safety culture from the long term point view. In this study, based on the site interviews to define the potential issues on organizational behavior for the safe operation and the survey on the level of safety culture of occupied workers are conducted. Survey results are quantified as a few indicators of nuclear safety by the statistical method and it can be simulated by the dynamic modeling as time goes on. Currently index and dynamic modeling are still being developed, however, results can be used to suggest the long term strategy which safety is clearly integrated into all activities in the nuclear organization

European passive plant program preliminary safety analyses to support system design

International Nuclear Information System (INIS)

Saiu, Gianfranco; Barucca, Luciana; King, K.J.

1999-01-01

In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

HACCP and water safety plans in Icelandic water supply: preliminary evaluation of experience.

Science.gov (United States)

Gunnarsdóttir, María J; Gissurarson, Loftur R

2008-09-01

Icelandic waterworks first began implementing hazard analysis and critical control points (HACCP) as a preventive approach for water safety management in 1997. Since then implementation has been ongoing and currently about 68% of the Icelandic population enjoy drinking water from waterworks with a water safety plan based on HACCP. Preliminary evaluation of the success of HACCP implementation was undertaken in association with some of the waterworks that had implemented HACCP. The evaluation revealed that compliance with drinking water quality standards improved considerably following the implementation of HACCP. In response to their findings, waterworks implemented a large number of corrective actions to improve water safety. The study revealed some limitations for some, but not all, waterworks in relation to inadequate external and internal auditing and a lack of oversight by health authorities. Future studies should entail a more comprehensive study of the experience with the use of HACCP with the purpose of developing tools to promote continuing success.

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

Safety and feasibility of transcranial direct current stimulation in pediatric hemiparesis: randomized controlled preliminary study.

Science.gov (United States)

Gillick, Bernadette T; Feyma, Tim; Menk, Jeremiah; Usset, Michelle; Vaith, Amy; Wood, Teddi Jean; Worthington, Rebecca; Krach, Linda E

2015-03-01

Transcranial direct current stimulation (tDCS) is a form of noninvasive brain stimulation that has shown improved adult stroke outcomes. Applying tDCS in children with congenital hemiparesis has not yet been explored. The primary objective of this study was to explore the safety and feasibility of single-session tDCS through an adverse events profile and symptom assessment within a double-blind, randomized placebo-controlled preliminary study in children with congenital hemiparesis. A secondary objective was to assess the stability of hand and cognitive function. A double-blind, randomized placebo-controlled pretest/posttest/follow-up study was conducted. The study was conducted in a university pediatric research laboratory. Thirteen children, ages 7 to 18 years, with congenital hemiparesis participated. Adverse events/safety assessment and hand function were measured. Participants were randomly assigned to either an intervention group or a control group, with safety and functional assessments at pretest, at posttest on the same day, and at a 1-week follow-up session. An intervention of 10 minutes of 0.7 mA tDCS was applied to bilateral primary motor cortices. The tDCS intervention was considered safe if there was no individual decline of 25% or group decline of 2 standard deviations for motor evoked potentials (MEPs) and behavioral data and no report of adverse events. No major adverse events were found, including no seizures. Two participants did not complete the study due to lack of MEP and discomfort. For the 11 participants who completed the study, group differences in MEPs and behavioral data did not exceed 2 standard deviations in those who received the tDCS (n=5) and those in the control group (n=6). The study was completed without the need for stopping per medical monitor and biostatisticial analysis. A limitation of the study was the small sample size, with data available for 11 participants. Based on the results of this study, tDCS appears to be safe

Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

International Nuclear Information System (INIS)

Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

2007-01-01

SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

Response to Absorber-Focus Coil Preliminary Safety Review Panel

International Nuclear Information System (INIS)

Barr, Giles; Baynham, Elwyn; Black, Edgar; Bradshaw, Tom; Cummings, Mary Anne; Green, Michael A.; Ishimoto, Shigeru; Ivanyushenkov, Yury; Lau, Wing; Zisman, Michael S.

2004-01-01

In this document we provide responses to the various issues raised in the report of the Preliminary Safety Review Panel (see http://mice.iit.edu/mnp/MICE0069.pdf). In some cases we have made design changes in response to the Panels suggestions. In other cases, we have chosen not to do so. In a few cases, we indicate our plans, although the tasks have not yet been completed. For simplicity, the responses are organized along the same lines as those of the Panel Report

Preliminary safety analysis of the Gorleben site

International Nuclear Information System (INIS)

Bracke, G.; Fischer-Appelt, K.

2014-01-01

The safety requirements governing the final disposal of heat-generating radioactive waste in Germany were implemented by the Federal Ministry of Environment, Natural Conservation and Nuclear Safety (BMU) in 2010. The Ministry considers as a fundamental objective the protection of man and the environment against the hazards of radioactive waste. Unreasonable burdens and obligation for future generations shall be avoided. The main safety principles are concentration and inclusion of radioactive and other pollutants in a containment-providing rock zone. Any release of radioactive nuclides may increase the risk for men and the environment only negligibly compared to natural radiation exposure. No intervention or maintenance work shall be necessary in the post-closure phase. Retrieval/recovery of the waste shall be possible up to 500 years after closure. The Gorleben salt dome has been discussed since the 1970's as a possible repository site for heat-generating radioactive waste in Germany. The objective of the project preliminary safety analysis of the Gorleben site (VSG) was to assess if repository concepts at the Gorleben site or other sites with a comparable geology could comply with these requirements based on currently available knowledge (Fischer-Appelt, 2013; Bracke, 2013). In addition to this it was assessed if methodological approaches can be used for a future site selection procedure and which technological and conceptual considerations can be transferred to other geological situations. The objective included the compilation and review of the available exploration data of the Gorleben site and on disposal in salt rock, the development of repository designs, and the identification of the needs for future R and D work and further site investigations. (authors)

Conclusion of the Preliminary Safety report for the LILW Repository on Trgovska Gora

International Nuclear Information System (INIS)

Lokner, V.; Levanat, I.; Schaller, A.; Kucar-Dragicevic, S.; Cerskov Klika, M.; Subasic, D.

2002-01-01

For more than a decade, APO d.o.o. has been engaged in preparations which might lead to establishment of a radioactive waste repository on Trgovska Gora, suitable for disposal of low and intermediate level waste (LILW) from the nuclear power plant Krsko. A recent product of theses activities is the preliminary safety assessment report (PSAR) for the proposed repository. In addition to an extensive overview of the repository project status, this preliminary SAR describes how the safety assessment methodology is used to demonstrate that a LILW facility will comply with radiological protection and safety requirements after the repository closure. LILW repository is designed to isolate waste from the environment for a couple hundred years in a reasonably efficient manner. It is generally not practicable to grant full waste containment throughout that period, because it suffices to demonstrate that radionuclide release and migration will remain below acceptable levels, which is achieved through safety assessment scenarios, modeling and calculations. However, with very limited repository specific data, safety assessment can only produce a conservative estimate of the upper bounds of potential exposures the repository could inflict. This PSAR arrives at such estimates in two different ways: (a) by simple bounding calculations and (b) through more sophisticated modeling and application of dedicated computer codes, but with similar conservative assumptions. Both approaches conservatively estimate that the highest potential dose to a nearby resident cannot significantly exceed the dose constraint of 0.2 mSv per year. Only in case of inadvertent intrusion into the near-surface disposal vault, much higher doses might be inflicted immediately after the planned institutional control of 250 years expires, but that can be prevented by a longer control period. Despite the preliminary and bounding style of the calculations, the PSAR has identified most important assumptions and

Mapping the nomological network of employee self-determined safety motivation: A preliminary measure in China.

Science.gov (United States)

Jiang, Li; Tetrick, Lois E

2016-09-01

The present study introduced a preliminary measure of employee safety motivation based on the definition of self-determination theory from Fleming (2012) research and validated the structure of self-determined safety motivation (SDSM) by surveying 375 employees in a Chinese high-risk organization. First, confirmatory factor analysis (CFA) was used to examine the factor structure of SDSM, and indices of five-factor model CFA met the requirements. Second, a nomological network was examined to provide evidence of the construct validity of SDSM. Beyond construct validity, the analysis also produced some interesting results concerning the relationship between leadership antecedents and safety motivation, and between safety motivation and safety behavior. Autonomous motivation was positively related to transformational leadership, negatively related to abusive supervision, and positively related to safety behavior. Controlled motivation with the exception of introjected regulation was negatively related to transformational leadership, positively related to abusive supervision, and negatively related to safety behavior. The unique role of introjected regulation and future research based on self-determination theory were discussed. Copyright © 2016 Elsevier Ltd. All rights reserved.

Preliminary safety evaluation for 241-C-106 waste retrieval, project W-320

International Nuclear Information System (INIS)

Conner, J.C.

1994-01-01

This document presents the Preliminary Safety Evaluation for Project W-320, Tank 241-C-106 Waste Retrieval Sluicing System (WRSS). The US DOE has been mandated to develop plans for response to safety issues associated with the waste storage tanks at the Hanford Site, and to report the progress of implementing those plans to Congress. The objectives of Project W-230 are to design, fabricate, develop, test, and operate a new retrieval system capable of removing a minimum of about 75% of the high-heat waste contained in C-106. It is anticipated that sluicing operations can remove enough waste to reduce the remaining radiogenic heat load to levels low enough to resolve the high-heat safety issue as well as allow closure of the tank safety issue

Preliminary Marine Safety Risk Assessment, Brandon Road Lock and Dam Invasive Species Control Measures

Science.gov (United States)

2016-12-01

Decision makers must include control-measure monitoring and emergency “interventions” to insure safety. The Coast Guard operational commanders...system” incorporates a travelling car on a rail above the barge-loading wharf to prevent loading personnel, cargo surveyors, or others from falling...to the Gulf of Mexico . As “Loopers”, they will have already transited the CSSC electric barriers. Preliminary Marine Safety Risk Assessment, BRLD

EPR safety. Consideration of the internal and external hazards in the safety studies

International Nuclear Information System (INIS)

Gueguin, H.

2008-04-01

The author presents the main points of the Preliminary Safety Report of EDF on the EPR reactor safety. It concerns the considerations of the internal (fire, flood, explosions, pipes failures) and external (earthquakes, airplane falls, explosions, exceptional natural disasters, extreme meteorological conditions) damages. It presents how the safety report takes into account the aggression. (A.L.B.)

Preliminary study on functional performance of compound type multistage safety injection tank

International Nuclear Information System (INIS)

Bae, Youngmin; Kim, Young In; Kim, Keung Koo

2015-01-01

Highlights: • Functional performance of compound type multistage safety injection tanks is studied. • Effects of key design parameters are scrutinized. • Distinctive flow features in compound type safety injection tanks are explored. - Abstract: A parametric study is carried out to evaluate the functional performance of a compound type multistage safety injection tank that would be considered one of the components for the passive safety injection systems in nuclear power plants. The effects of key design parameters such as the initial volume fraction and charging pressure of gas, tank elevation, vertical location of a sparger, resistance coefficient, and operating condition on the injection flow rate are scrutinized along with a discussion of the relevant flow features. The obtained results indicate that the compound type multistage safety injection tank can effectively control the injection flow rate in a passive manner, by switching the driving force for the safety injection from gas pressure to gravity during the refill and reflood phases, respectively

Hydrogen Gas Retention and Release from WTP Vessels: Summary of Preliminary Studies

Energy Technology Data Exchange (ETDEWEB)

Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bontha, Jagannadha R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mahoney, Lenna A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rassat, Scot D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Boeringa, Gregory K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buchmiller, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chun, Jaehun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Karri, Naveen K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Li, Huidong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tran, Diana N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-07-01

The Hanford Waste Treatment and Immobilization Plant (WTP) is currently being designed and constructed to pretreat and vitrify a large portion of the waste in the 177 underground waste storage tanks at the Hanford Site. A number of technical issues related to the design of the pretreatment facility (PTF) of the WTP have been identified. These issues must be resolved prior to the U.S. Department of Energy (DOE) Office of River Protection (ORP) reaching a decision to proceed with engineering, procurement, and construction activities for the PTF. One of the issues is Technical Issue T1 - Hydrogen Gas Release from Vessels (hereafter referred to as T1). The focus of T1 is identifying controls for hydrogen release and completing any testing required to close the technical issue. In advance of selecting specific controls for hydrogen gas safety, a number of preliminary technical studies were initiated to support anticipated future testing and to improve the understanding of hydrogen gas generation, retention, and release within PTF vessels. These activities supported the development of a plan defining an overall strategy and approach for addressing T1 and achieving technical endpoints identified for T1. Preliminary studies also supported the development of a test plan for conducting testing and analysis to support closing T1. Both of these plans were developed in advance of selecting specific controls, and in the course of working on T1 it was decided that the testing and analysis identified in the test plan were not immediately needed. However, planning activities and preliminary studies led to significant technical progress in a number of areas. This report summarizes the progress to date from the preliminary technical studies. The technical results in this report should not be used for WTP design or safety and hazards analyses and technical results are marked with the following statement: “Preliminary Technical Results for Planning – Not to be used for WTP Design

Preliminary safety evaluation (PSE) for Sodium Storage Facility at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Bowman, B.R.

1994-01-01

This evaluation was performed for the Sodium Storage Facility (SSF) which will be constructed at the Fast Flux Test Facility (FFTF) in the area adjacent to the South and West Dump Heat Exchanger (DHX) pits. The purpose of the facility is to allow unloading the sodium from the FFTF plant tanks and piping. The significant conclusion of this Preliminary Safety Evaluation (PSE) is that the only Safety Class 2 components are the four sodium storage tanks and their foundations. The building, because of its imminent risk to the tanks under an earthquake or high winds, will be Safety Class 3/2, which means the building has a Safety Class 3 function with the Safety Class 2 loads of seismic and wind factored into the design

[Preliminary results of an open-label observational study evaluating the efficacy and safety of Prolia used in women with postmenopausal osteoporosis].

Science.gov (United States)

Ershova, O B; Lesniak, O M; Belova, K Iu; Nazarova, A V; Manovitskaia, A V; Musaeva, T M; Musraev, R M; Nurlygaianov, R Z; Rozhinskaia, L Ia; Skripnikova, I A; Toroptsova, N V

2014-01-01

To evaluate the efficacy and safety of Denosumab (Prolia), a first-line osteoporosis (OP) medication that is a fully human monoclonal antibody to the receptor activator of nuclear factor xB ligand (RANKL), within an open-label observational study. Patients aged 50 years or older with postmenopausal OP, who were treated with Prolia in clinical practice, were examined. The concentrations of the bone resorption (BR) marker of C-terminal telopeptide and other laboratory indicators (total serum calcium, total alkaline phosphatase, and creatinine) were measured following 3 months. Adverse drug reactions were recorded. Three months after initiation of the investigation, there was a significant decrease in the BR marker C-terminal telopeptide (by 89%; p<0.0001). There were rare adverse reactions: hypocalcemia in 3 (5.9%) patients, arthralgias in 2 (3.9%), and eczema in 1 (1.9%). There were neither serious adverse events nor study withdrawal cases. The preliminary results of the open-label study of Prolia in postmenopausal OP suggest that the significantly lower BR activity determines the efficacy of this drug and its high safety.

Customer Relationship Management System in Occupational Safety & Health Companies: Research on Practice and Preliminary Design Solution

Directory of Open Access Journals (Sweden)

Robert Fabac

2011-10-01

Full Text Available One of the most prominent contemporary trends in formation of companies is the approach to development of a customer-oriented company. In this matter, various versions related to the intensity of this orientation are differentiated. Customer relationship management (CRM system is a well-known concept, and its practice is being studied and improved in connection to various sectors. Companies providing services of occupational safety and health (OHS mainly cooperate with a large number of customers and the quality of this cooperation largely affects the occupational safety and health of employees. Therefore, it is of both scientific and wider social interest to study and improve the relationship of these companies with their customers. This paper investigates the practice of applying CRM in Croatian OHS companies. It identifies the existing conditions and suggests possible improvements in the practice of CRM, based on experts’ assessments using analytic hierarchy process evaluation. Universal preliminary design was created as a framework concept for the formation of a typical customer-oriented OHS services company. Preliminary design includes a structural view, which provides more details through system diagrams, and an illustration of main cooperation processes of a company with its customer.

Developing a disaster education program for community safety and resilience: The preliminary phase

Science.gov (United States)

Nifa, Faizatul Akmar Abdul; Abbas, Sharima Ruwaida; Lin, Chong Khai; Othman, Siti Norezam

2017-10-01

Resilience encompasses both the principles of preparedness and reaction within the dynamic systems and focuses responses on bridging the gap between pre-disaster activities and post-disaster intervention and among structural/non-structural mitigation. Central to this concept is the ability of the affected communities to recover their livelihood and inculcating necessary safety practices during the disaster and after the disaster strikes. While these ability and practices are important to improve the community safety and resilience, such factors will not be effective unless the awareness is present among the community. There have been studies conducted highlighting the role of education in providing awareness for disaster safety and resilience from a very young age. However for Malaysia, these area of research has not been fully explored and developed based on the specific situational and geographical factors of high-risk flood disaster locations. This paper explores the importance of disaster education program in Malaysia and develops into preliminary research project which primary aim is to design a flood disaster education pilot program in Kampung Karangan Primary School, Kelantan, Malaysia.

Development of safety analysis technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

1999-03-01

Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

Preliminary study to improve the performance of SCWR-M during loss-of-flow accident

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Sun, C.; Wang, Z.D.; Chai, X.; Xiong, J.B.; Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

2016-10-15

Highlights: • Validation of the ATHLET-SC code to the safety analysis for SCWR. • Loss of flow accident analysis for SCWR-M is performed. • The passive design parameter is optimized. • The optimized SCWR-M design shows a better safety performance. - Abstract: The SCWR-M is one of the conceptual core designs with mixed neutron spectrum (fast and thermal), which is developed at Shanghai Jiao Tong University. Some preliminary calculations of this new conceptual SCWR indicate the SCWR-M system gets better safety characteristics compared to other single spectrum supercritical water cooled reactors. Loss of flow accident (LOFA) is of particular importance among the abnormal events and accidents for SCWR-M. In order to perform the preliminary study to improve the current SCWR-M safety design, this paper presents the validation results of the ATHLET-SC code and optimization work for safety system design parameters of the ICS, ACC, GDCS based on LOFA analysis. The better performance of the optimized design parameters are demonstrated by comparison with the previous design.

Preliminary post-closure safety assessment of repository concepts for low level radioactive waste at the Bruce Site, Ontario

International Nuclear Information System (INIS)

Little, R.H.; Penfold, J.S.S.; Egan, M.J.; Leung, H.

2005-01-01

The preliminary post-closure safety assessment of permanent repository concepts for low-level radioactive waste (LLW) at the Ontario Power Generation (OPG) Bruce Site is described. The study considered the disposal of both short and long-lived LLW. Four geotechnically feasible repository concepts were considered (two near-surface and two deep repositories). An approach consistent with best international practice was used to provide a reasoned and comprehensive analysis of post-closure impacts of the repository concepts. The results demonstrated that the deep repository concepts in shale and in limestone, and the surface repository concept on sand should meet radiological protection criteria. For the surface repository concept on glacial till, it appears that increased engineering such as grouting of waste and voids should be considered to meet the relevant dose constraint. Should the project to develop a permanent repository for LLW proceed, it is expected that this preliminary safety assessment would need to be updated to take account of future site-specific investigations and design updates. (author)

Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

International Nuclear Information System (INIS)

Herborn, D.I.

1991-10-01

The requirements for Westinghouse Hanford independent review of the Preliminary Safety Analysis Report (PSAR) are contained in Section 1.0, Subsection 4.3 of WCH-CM-4-46. Specifically, this manual requires the following: (1) Formal functional reviews of the HWVP PSAR by the future operating organization (HWVP Operations), and the independent review organizations (HWVP and Environmental Safety Assurance, Environmental Assurance, and Quality Assurance); and (2) Review and approval of the HWVP PSAR by the Tank Waste Disposal (TWD) Subcouncil of the Safety and Environmental Advisory Council (SEAC), which provides independent advice to the Westinghouse Hanford President and executives on matters of safety and environmental protection. 7 refs

Preliminary Safety Analysis Report for the Transuranic Storage Area Retrieval Enclosure at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

1993-03-01

This Transuranic Storage Area Retrieval Enclosure Preliminary Safety Analysis Report was completed as required by DOE Order 5480.23. The purpose of this document is to construct a safety basis that supports the design and permits construction of the facility. The facility has been designed to the requirements of a Radioactive Solid Waste Facility presented in DOE Order 6430.1A

Posiva's application for a decision in principle concerning a disposal facility for spent nuclear fuel. STUK's statement and preliminary safety appraisal

International Nuclear Information System (INIS)

Ruokola, E.

2000-03-01

In May 1999, Posiva Ltd submitted to the Government an application, pursuant to the Nuclear Energy Act, for a Decision in Principle on a disposal facility for spent nuclear fuel from the Finnish nuclear power plants. The Ministry of Trade and Industry requested the Radiation and Nuclear Safety Authority (STUK) to draw up a preliminary safety appraisal concerning the proposed disposal facility. In the beginning of this report, STUK's statement to the Ministry and Industry concerning the proposed disposal facility is given. In that statement, STUK concludes that the Decision in Principle is currently justified from the standpoint of safety. The statement is followed by a safety appraisal, where STUK deems, how the proposed disposal concept, site and facility comply with the safety requirements included in the Government's Decision (478/1999). STUK's preliminary safety appraisal was supported by contributions from a number of outside experts. A collective opinion by an international group of ten distinguished experts is appended to this report. (orig.)

Feasibility, safety, acceptability, and preliminary efficacy of measurement-based care depression treatment for HIV patients in Bamenda, Cameroon.

Science.gov (United States)

Pence, Brian W; Gaynes, Bradley N; Atashili, Julius; O'Donnell, Julie K; Kats, Dmitry; Whetten, Kathryn; Njamnshi, Alfred K; Mbu, Tabenyang; Kefie, Charles; Asanji, Shantal; Ndumbe, Peter

2014-06-01

Depression affects 18-30 % of HIV-infected patients in Africa and is associated with greater stigma, lower antiretroviral adherence, and faster disease progression. However, the region's health system capacity to effectively identify and treat depression is limited. Task-shifting models may help address this large mental health treatment gap. Measurement-Based Care (MBC) is a task-shifting model in which a Depression Care Manager guides a non-psychiatric (e.g., HIV) provider in prescribing and managing antidepressant treatment. We adapted MBC for depressed HIV-infected patients in Cameroon and completed a pilot study to assess feasibility, safety, acceptability, and preliminary efficacy. We enrolled 55 participants; all started amitriptyline 25-50 mg daily at baseline. By 12 weeks, most remained at 50 mg daily (range 25-125 mg). Median (interquartile range) PHQ-9 depressive severity scores declined from 13 (12-16) (baseline) to 2 (0-3) (week 12); 87 % achieved depression remission (PHQ-9 feasibility, safety, acceptability, and preliminary efficacy in this uncontrolled pilot study. Further research should assess whether MBC could improve adherence and HIV outcomes in this setting.

Preliminary safety evaluation of an aircraft impact on a near-surface radioactive waste repository

Energy Technology Data Exchange (ETDEWEB)

Lo Frano, R.; Forasassi, G.; Pugliese, G. [Department of Industrial and Civil Engineering (DICI), University of Pisa, Pisa (Italy)

2013-07-01

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a 'beyond design basis' event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a 'robust' target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircraft, even in presence of ongoing concrete progressive failure (some penetration and spalling of the

Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

International Nuclear Information System (INIS)

Herborn, D.I.

1993-11-01

Westinghouse Hanford Company (WHC) is the Integrating Contractor for the Hanford Waste Vitrification Plant (HWVP) Project, and as such is responsible for preparation of the HWVP Preliminary Safety Analysis Report (PSAR). The HWVP PSAR was prepared pursuant to the requirements for safety analyses contained in US Department of Energy (DOE) Orders 4700.1, Project Management System (DOE 1987); 5480.5, Safety of Nuclear Facilities (DOE 1986a); 5481.lB, Safety Analysis and Review System (DOE 1986b) which was superseded by DOE order 5480-23, Nuclear Safety Analysis Reports, for nuclear facilities effective April 30, 1992 (DOE 1992); and 6430.lA, General Design Criteria (DOE 1989). The WHC procedures that, in large part, implement these DOE requirements are contained in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals

Obtaining Valid Safety Data for Software Safety Measurement and Process Improvement

Science.gov (United States)

Basili, Victor r.; Zelkowitz, Marvin V.; Layman, Lucas; Dangle, Kathleen; Diep, Madeline

2010-01-01

We report on a preliminary case study to examine software safety risk in the early design phase of the NASA Constellation spaceflight program. Our goal is to provide NASA quality assurance managers with information regarding the ongoing state of software safety across the program. We examined 154 hazard reports created during the preliminary design phase of three major flight hardware systems within the Constellation program. Our purpose was two-fold: 1) to quantify the relative importance of software with respect to system safety; and 2) to identify potential risks due to incorrect application of the safety process, deficiencies in the safety process, or the lack of a defined process. One early outcome of this work was to show that there are structural deficiencies in collecting valid safety data that make software safety different from hardware safety. In our conclusions we present some of these deficiencies.

Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

Energy Technology Data Exchange (ETDEWEB)

Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

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

International Nuclear Information System (INIS)

1999-11-01

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

Preliminary safety analysis of molten salt breeder reactor

International Nuclear Information System (INIS)

Cheng Maosong; Dai Zhimin

2013-01-01

Background: The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. ORNL finished the conceptual design of Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). Purpose: We attempt to implement the preliminary safety analysis of MSBR in order to provide a reference for the design and optimization of MSBR in the future. Methods: According to the conceptual design of MSBR, a model of safety analysis using point kinetics coupled with the simplified heat transfer mechanism is presented. The model is applied to simulate the transient phenomena of MSBR initiated by an abnormal step reactivity addition and an abnormal ramp reactivity addition at full-power equilibrium condition. Results: The thermal power in the core increases rapidly at the beginning and is accompanied by a rise of the fuel and graphite temperatures after 100, 300, 500 and 600 pcm reactivity addition. The maximum outlet temperature of the fuel in the core is at 1250℃ in 500 pcm reactivity addition, but up to 1350℃ in 600 pcm reactivity addition. The maximum of the power and the temperature are delayed and lower in the ramp reactivity addition rather than in the step reactivity addition. Conclusions: Based on the results, when the reactivity inserted is less than 500 pcm in maximum at full power equilibrium condition, the structural material in Hastelloy-N is not melted and can keep integrity without external control action. And it is necessary to try to avoid inserting a reactivity at short time. (authors)

Preliminary investigation of interconnected systems interactions for the safety injection system of Indian Point-3

International Nuclear Information System (INIS)

Alesso, H.P.; Lappa, D.A.; Smith, C.F.; Sacks, I.J.

1983-01-01

The rich diversity of ideas and techniques for analyzing interconnected systems interaction has presented the NRC with the problem of identifying methods appropriate for their own review and audit. This report presents the findings of a preliminary study using the Digraph Matrix Analysis method to evaluate interconnected systems interactions for the safety injection system of Indian Point-3. The analysis effort in this study was subjected to NRC constraints regarding the use of Boolean logic, the construction of simplified plant representations or maps, and the development of heuristic measures as specified by the NRC. The map and heuristic measures were found to be an unsuccessful approach. However, from the effort to model and analyze the Indian Point-3 safety injection system, including Boolean logic in the model, singleton and doubleton cut-sets were identified. It is recommended that efforts excluding Boolean logic and utilizing the NRC heuristic measures not be pursed further and that the Digraph Matrix approach (or other comparable risk assessment technique) with Boolean logic included to conduct the audit of the Indian Point-3 systems interaction study

A study of software safety analysis system for safety-critical software

International Nuclear Information System (INIS)

Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

2004-01-01

The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

Keeping rail on track: preliminary findings on safety culture in Australian rail.

Science.gov (United States)

Blewett, Verna; Rainbird, Sophia; Dorrian, Jill; Paterson, Jessica; Cattani, Marcus

2012-01-01

'Safety culture' is identified in the literature as a critical element of healthy and safe workplaces. How can rail organizations ensure that consistently effective work health and safety cultures are maintained across the diversity of their operations? This paper reports on research that is currently underway in the Australian rail industry aimed at producing a Model of Best Practice in Safety Culture for the industry. Located in rail organizations dedicated to the mining industry as well as urban rail and national freight operations, the research examines the constructs of organizational culture that impact on the development and maintenance of healthy and safe workplaces. The research uses a multi-method approach incorporating quantitative (survey) and qualitative (focus groups, interviews and document analysis) methods along with a participative process to identify interventions to improve the organization and develop plans for their implementation. The research uses as its analytical framework the 10 Platinum Rules, from the findings of earlier research in the New South Wales (Australia) mining industry, Digging Deeper. Data collection is underway at the time of writing and preliminary findings are presented at this stage. The research method may be adapted for use as a form of organizational review of safety and health in organizational culture.

Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

International Nuclear Information System (INIS)

Lydell, B.

1998-03-01

This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood

Preliminary safety evaluation for the Laxemar subarea. Based on data and site descriptions after the initial site investigation stage

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan [JA Streamflow AB, Aelvsjoe (Sweden)

2006-03-15

The main objectives of this Preliminary Safety Evaluation (PSE) of the Laxemar subarea have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB in 2000. These criteria both concern properties of the site judged to be necessary for safety and engineering (requirements) and properties judged to be beneficial (preferences). The findings are then evaluated in order to provide feedback to continued investigations and design work. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The latter is eventually done in coming Safety Assessments. This preliminary safety evaluation shows that, according to existing data, the Laxemar subarea meets all safety requirements. The evaluation also shows that the Laxemar subarea meets most of the safety preferences, but for some aspects of the site description further reduction of the uncertainties would enhance the safety case. Despite the stated concerns, there is no reason, from a safety point of view, not to continue the Site Investigations at the Laxemar subarea. There are uncertainties to resolve and the safety would eventually need to be verified through a proper safety assessment. Only some of the uncertainties noted in the Site Descriptive Model have safety implications and need further resolution for this reason. Furthermore, uncertainties may need resolving for other reasons, such as giving an adequate assurance of site understanding or assisting in optimising design. Notably, there are questions about the

Preliminary safety evaluation for the Laxemar subarea. Based on data and site descriptions after the initial site investigation stage

International Nuclear Information System (INIS)

Andersson, Johan

2006-03-01

The main objectives of this Preliminary Safety Evaluation (PSE) of the Laxemar subarea have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB in 2000. These criteria both concern properties of the site judged to be necessary for safety and engineering (requirements) and properties judged to be beneficial (preferences). The findings are then evaluated in order to provide feedback to continued investigations and design work. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The latter is eventually done in coming Safety Assessments. This preliminary safety evaluation shows that, according to existing data, the Laxemar subarea meets all safety requirements. The evaluation also shows that the Laxemar subarea meets most of the safety preferences, but for some aspects of the site description further reduction of the uncertainties would enhance the safety case. Despite the stated concerns, there is no reason, from a safety point of view, not to continue the Site Investigations at the Laxemar subarea. There are uncertainties to resolve and the safety would eventually need to be verified through a proper safety assessment. Only some of the uncertainties noted in the Site Descriptive Model have safety implications and need further resolution for this reason. Furthermore, uncertainties may need resolving for other reasons, such as giving an adequate assurance of site understanding or assisting in optimising design. Notably, there are questions about the

Preliminary study: Formaldehyde exposure in laboratories of Sharjah university in UAE

OpenAIRE

Ahmed, Hafiz Omer

2011-01-01

Objectives : Laboratory technicians, students, and instructors are at high risk, because they deal with chemicals including formaldehyde. Thus, this preliminary study was conducted to measure the concentration of formaldehyde in the laboratories of the University of Sharjah in UAE. Materials and Methods: Thirty-two air samples were collected and analyzed for formaldehyde using National Institute for Occupational Safety and Health (NIOSH) method 3500. In this method, formaldehyde reacts with c...

Deep Brain Stimulation in Huntington’s Disease—Preliminary Evidence on Pathophysiology, Efficacy and Safety

Directory of Open Access Journals (Sweden)

Lars Wojtecki

2016-08-01

Full Text Available Huntington’s disease (HD is one of the most disabling degenerative movement disorders, as it not only affects the motor system but also leads to cognitive disabilities and psychiatric symptoms. Deep brain stimulation (DBS of the pallidum is a promising symptomatic treatment targeting the core motor symptom: chorea. This article gives an overview of preliminary evidence on pathophysiology, safety and efficacy of DBS in HD.

Geoscientific long-term prognosis. Preliminary safety analysis for the site Gorleben

International Nuclear Information System (INIS)

Mrugalla, Sabine

2011-07-01

The preliminary safety analysis of the site Gorleben includes the following chapters: (1) Introduction; (2) Aim and content of the geoscientific long-term prognosis for the site Gorleben; (3) Boundary conditions at the site Gorleben: climate; geomorphology; overlying rocks and adjoining rocks; hydrogeology; salt deposit Gorleben. (4) Probable future geological developments at the site Gorleben: supraregional developments with effects on the site Gorleben; glacial period developments; developments of the geomorphology, overlying and adjoining rocks; future developments of the hydrological systems at the site Gorleben; future saliniferous specific developments of the salt deposit Gorleben. (5) Commentary on the unlikely or excludable developments of the site Gorleben.

Comprehensive development plans for the low- and intermediate-level radioactive waste disposal facility in Korea and preliminary safety assessment

Energy Technology Data Exchange (ETDEWEB)

Jung, Kang Il; Kim, Jin Hyeong; Kwon, Mi Jin; Jeong, Mi Seon; Hong, Sung Wook; Park, Jin Beak [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

The disposal facility in Gyeongju is planning to dispose of 800,000 packages of low- and intermediate- level radioactive waste. This facility will be developed as a complex disposal facility that has various types of disposal facilities and accompanying management. In this study, based on the comprehensive development plan of the disposal facility, a preliminary post-closure safety assessment is performed to predict the phase development of the total capacity for the 800,000 packages to be disposed of at the site. The results for each scenario meet the performance target of the disposal facility. The assessment revealed that there is a significant impact of the inventory of intermediate-level radionuclide waste on the safety evaluation. Due to this finding, we introduce a disposal limit value for intermediate-level radioactive waste. With stepwise development of safety case, this development plan will increase the safety of disposal facilities by reducing uncertainties within the future development of the underground silo disposal facilities.

A study on methodologies for assessing safety critical network's risk impact on Nuclear Power Plant

International Nuclear Information System (INIS)

Lim, T. J.; Lee, H. J.; Park, S. K.; Seo, S. J.

2006-08-01

The objectives of this project is to investigate and study existing reliability analysis techniques for communication networks in order to develop reliability analysis models for Nuclear Power Plant's safety-critical networks. It is necessary to make a comprehensive survey of current methodologies for communication network reliability. Major outputs of the first year study are design characteristics of safety-critical communication networks, efficient algorithms for quantifying reliability of communication networks, and preliminary models for assessing reliability of safety-critical communication networks

Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

Energy Technology Data Exchange (ETDEWEB)

Lydell, B. [RSA Technologies, Vista, CA (United States)

1998-03-01

This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood. 7 refs, 4 tabs, 3 figs. Also available at the SKI Home page: //www.ski.se.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-11-01

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

Preliminary Evaluation of an Aviation Safety Thesaurus' Utility for Enhancing Automated Processing of Incident Reports

Science.gov (United States)

Barrientos, Francesca; Castle, Joseph; McIntosh, Dawn; Srivastava, Ashok

2007-01-01

This document presents a preliminary evaluation the utility of the FAA Safety Analytics Thesaurus (SAT) utility in enhancing automated document processing applications under development at NASA Ames Research Center (ARC). Current development efforts at ARC are described, including overviews of the statistical machine learning techniques that have been investigated. An analysis of opportunities for applying thesaurus knowledge to improving algorithm performance is then presented.

Operating performance and environmental and safety risks: A preliminary comparison of majors and independents

International Nuclear Information System (INIS)

Pulsipher, A.G.; Iledare, W.O.; Baumann, R.H.; Mesyanzhinov, D.

1995-01-01

The objective is to compare the safety and environmental records of oil and gas companies operating on the OCS in the Gulf of Mexico over the past decade. The reason for doing so is to help inform public sector policy-makers and private sector decision-makers about the potential safety and environmental risks associated with the expected increased presence of smaller independents in the domestic oil and gas industry in general and on the federal OCS in particular. The preliminary conclusion is that although independents have had a modestly high incidence of fires and explosions than the majors, the difference is not significant statistically and is largely attributable to a few ''bad actors'' rather than demonstrably poorer practice by the group as a whole

Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant; Resultados preliminares mas significativos del analysis probabilista de seguridad de la Central Nuclear de Juragua

Energy Technology Data Exchange (ETDEWEB)

Perdomo, Manuel [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

1995-12-31

Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential `weak points` at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs.

Preliminary neutronic study on Pu-based OTTO cycle pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Setiadipura, Topan; Zuhair [National Nuclear Energy Agency of Indonesia (BATAN), Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Irwanto, Dwi [Bandung Institute of Technology (ITB), Bandung (Indonesia). Nuclear Physics and Biophysics Research Group

2017-12-15

The neutron physics characteristic of Pebble Bed Reactor (PBR) allows a better incineration of plutonium (Pu). An optimized design of simple PBR might give a symbiotic solution of providing a safe energy source, effective fuel utilization shown by a higher burnup value, and incineration of Pu stockpiles. This study perform a preliminary neutronic design study of a 200 MWt Once Through Then Out (OTTO) cycle PBR with Pu-based fuel. The safety criteria of the design were represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. In this preliminary phase, the parametric survey is limited to the heavy metal (HM) loading per pebble and the average axial speed of the fuel. An optimum high burnup of 419.7 MWd/kg-HM was achieved in this study. This optimum design uses a HM loading of 2.5 g/pebble with average axial fuel velocity 0.5 cm/day.

ADME studies and preliminary safety pharmacology of LDT5, a lead compound for the treatment of benign prostatic hyperplasia

Directory of Open Access Journals (Sweden)

F. Noël

Full Text Available This study aimed to estimate the absorption, distribution, metabolism and excretion (ADME properties and safety of LDT5, a lead compound for oral treatment of benign prostatic hyperplasia that has previously been characterized as a multi-target antagonist of α1A-, α1D-adrenoceptors and 5-HT1A receptors. The preclinical characterization of this compound comprised the evaluation of its in vitro properties, including plasma, microsomal and hepatocytes stability, cytochrome P450 metabolism and inhibition, plasma protein binding, and permeability using MDCK-MDR1 cells. De-risking and preliminary safety pharmacology assays were performed through screening of 44 off-target receptors and in vivo tests in mice (rota-rod and single dose toxicity. LDT5 is stable in rat and human plasma, human liver microsomes and hepatocytes, but unstable in rat liver microsomes and hepatocytes (half-life of 11 min. LDT5 is highly permeable across the MDCK-MDR1 monolayer (Papp ∼32×10-6 cm/s, indicating good intestinal absorption and putative brain penetration. LDT5 is not extensively protein-bound and is a substrate of human CYP2D6 and CYP2C19 but not of CYP3A4 (half-life >60 min, and did not significantly influence the activities of any of the human cytochrome P450 isoforms screened. LDT5 was considered safe albeit new studies are necessary to rule out putative central adverse effects through D2, 5-HT1A and 5-HT2B receptors, after chronic use. This work highlights the drug-likeness properties of LDT5 and supports its further preclinical development.

Study of fast reactor safety test facilities. Preliminary report

International Nuclear Information System (INIS)

Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

1975-05-01

Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

A preliminary study on safety stock placement in capacitated supply chains

NARCIS (Netherlands)

Sitompul, Carles; Aghezzaf, El Houssaine; Chen, Huan; Dullaert, Wout

2006-01-01

The issue of safety stock placement is one of the challenging problems in the area of supply chain design. Safety stocks play a significant role in supply chains since they assure the service level and the smoothness of the flow of materials throughout the chain. Some special models of the problem

A PRELIMINARY STUDY ON SAFETY STOCK PLACEMENT IN CAPACITATED SUPPLY CHAINS

NARCIS (Netherlands)

Sitompul, Carles; Aghezzaf, El Houssaine; Chen, Huan; Dullaert, Wout; Dolgui, A.; Morel, G.; Pereira, C.E.

2006-01-01

Abstract The issue of safety stock placement is one of the challenging problems in the area of supply chain design. Safety stocks play a significant role in supply chains since they assure the service level and the smoothness of the flow of materials throughout the chain. Some special models of the

The study on length and diameter ratio of nail as preliminary design for slope stabilization

Science.gov (United States)

Gunawan, Indra; Silmi Surjandari, Niken; Muslih Purwana, Yusep

2017-11-01

Soil nailing technology has been widely applied in practice for reinforced slope. The number of studies for the effective design of nail-reinforced slopes has also increased. However, most of the previous study was focused on a safety factor of the slope; the ratio of length and diameter itself has likely never been studied before. The aim of this study is to relate the length and diameter ratio of the nail with the safety factor of the 20 m height of sand slope in the various angle of friction and steepness of the slope. Simplified Bishop method was utilized to analyze the safety factor of the slope. This study is using data simulation to calculate the safety factor of the slope with soil nailing reinforcement. The results indicate that safety factor of slope stability increases with the increase of length and diameter ratio of the nail. At any angle of friction and steepness of the slope, certain effective length and diameter ratio was obtain. These results may be considered as a preliminary design for slope stabilization.

Optimization study and preliminary design for Latina NPP early core retrieval and reactor dismantling

International Nuclear Information System (INIS)

Macci, E.; Zirpolo, S.; Imparato, A.; Cacace, A.; Parry, D.; Walkden, P.

2002-01-01

In June 2000, an agreement was established between Sogin and BNFL to enable the two companies to co-operate, using their specific experiences in the decommissioning field, for the benefit of projects in Italy, the United Kingdom and for third markets. A decommissioning strategy for the Latina NPP was initially developed in a Phase 1 Study which produced a conceptual design for the decommissioning of the reactor. This study was completed in June 2000. Since then, a second study has been completed, which has further developed the strategy and produced preliminary designs for the early dismantling of the core and reactor building at Latina. The engineering and safety data were produced in order to support Sogin in the preparation of a safety case for plant decommissioning. This safety case was submitted to the Italian Regulator, ANPA, in February 2002. (author)

Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

Energy Technology Data Exchange (ETDEWEB)

Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

2006-01-15

This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

Preliminary Development of Regulatory PSA Models for SFR

International Nuclear Information System (INIS)

Choi, Yong Won; Shin, Andong; Bae, Moohoon; Suh, Namduk; Lee, Yong Suk

2013-01-01

Well developed PRA methodology exists for LWR (Light Water Reactor) and PHWR (Pressurized Heavy Water Reactor). Since KAERI is developing a prototype SFR targeting to apply for a license by 2017, KINS needs to have a PRA models to assess the safety of this prototype reactor. The purpose of this study is to develop the regulatory PSA models for the independent verification of the SFR safety. Since the design of the prototype SFR is not mature yet, we have tried to develop the preliminary models based on the design data of KAERI's previous SFR design. In this study, the preliminary initiating events of level 1 internal event for SFR were selected through reviews of existing PRA (LWR, PRISM, ASTRID and KALIMER-600) models. Then, the event tree for each selected initiating event was developed. The regulatory PRA models of SFR developed are preliminary in a sense, because the prototype SFR design is not mature and provided yet. Still it might be utilized for the forthcoming licensing review in assessing the risk of safety issues and the configuration control of the design

Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

2017-07-01

In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

Scrambler therapy efficacy and safety for neuropathic pain correlated with chemotherapy-induced peripheral neuropathy in adolescents: A preliminary study.

Science.gov (United States)

Tomasello, Caterina; Pinto, Rita Maria; Mennini, Chiara; Conicella, Elena; Stoppa, Francesca; Raucci, Umberto

2018-04-06

Chemotherapy-induced peripheral neuropathy (CIPN) is a common side effect of chemotherapy, in need of effective treatment. Preliminary data support the efficacy of scrambler therapy (ST), a noninvasive cutaneous electrostimulation device, in adults with CIPN. We test the efficacy, safety, and durability of ST for neuropathic pain in adolescents with CIPN. We studied nine pediatric patients with cancer and CIPN who received ST for pain control. Each patient received 45-min daily sessions for 10 consecutive days as a first step, but some of them required additional treatment. Pain significantly improved comparing Numeric Rate Scale after 10 days of ST (9.22 ± 0.83 vs. 2.33 ± 2.34; P < 0.001) and at the end of the optimized cycle (EOC) (9.22 ± 0.83 vs. 0.11 ± 0.33, P < 0.001). The improvement in quality of life was significantly reached on pain interference with general activity (8.67 ± 1.66 vs. 3.33 ± 2.12, P < 0.0001), mood (8.33 ± 3.32 vs. 2.78 ± 2.82, P < 0.0005), walking ability (10.00 vs. 2.78 ± 1.22, P < 0.0001), sleep (7.56 ± 2.24 vs. 2.67 ± 1.41, P < 0.001), and relations with people (7.89 ± 2.03 vs. 2.11 ± 2.03, P < 0.0002; Lansky score 26.7 ± 13.2 vs. 10 days of ST 57.8 ± 13.9, P < 0.001; 26.7 ± 13.2 vs. EOC 71.1 ± 16.2, P < 0.001). Based on these preliminary data, ST could be a good choice for adolescents with CIPN for whom pain control is difficult. ST caused total relief or dramatic reduction in CIPN pain and an improvement in quality of life, durable in follow-up. It caused no detected side effects, and can be retrained successfully. Further larger studies should be performed to confirm our promising preliminary data in pediatric patients with cancer. © 2018 Wiley Periodicals, Inc.

Feasibility and Safety of a Powered Exoskeleton for Assisted Walking for Persons With Multiple Sclerosis: A Single-Group Preliminary Study.

Science.gov (United States)

Kozlowski, Allan J; Fabian, Michelle; Lad, Dipan; Delgado, Andrew D

2017-07-01

To examine the feasibility, safety, and secondary benefit potential of exoskeleton-assisted walking with one device for persons with multiple sclerosis (MS). Single-group longitudinal preliminary study with 8-week baseline, 8-week intervention, and 4-week follow-up. Outpatient MS clinic, tertiary care hospital. Participants (N=13; age range, 38-62y) were mostly women with Expanded Disability Status Scale scores ranging from 5.5 to 7.0. Exoskeleton-assisted walk training. Primary outcomes were accessibility (enrollment/screen pass), tolerability (completion/dropout), learnability (time to event for standing, walking, and sitting with little or no assistance), acceptability (satisfaction on the device subscale of the Quebec User Evaluation of Satisfaction with Assistive Technology version 2), and safety (event rates standardized to person-time exposure in the powered exoskeleton). Secondary outcomes were walking without the device (timed 25-foot walk test and 6-minute walk test distance), spasticity (Modified Ashworth Scale), and health-related quality of life (Patient-Reported Outcomes Measurement and Information System pain interference and Quality of Life in Neurological Conditions fatigue, sleep disturbance, depression, and positive affect and well-being). The device was accessible to 11 and tolerated by 5 participants. Learnability was moderate, with 5 to 15 sessions required to walk with minimal assistance. Safety was good; the highest adverse event rate was for skin issues at 151 per 1000 hours' exposure. Acceptability ranged from not very satisfied to very satisfied. Participants who walked routinely improved qualitatively on sitting, standing, or walking posture. Two participants improved and 2 worsened on ≥1 quality of life domain. The pattern of spasticity scores may indicate potential benefit. The device appeared feasible and safe for about a third of our sample, for whom routine exoskeleton-assisted walking may offer secondary benefits. Copyright �

Preliminary design studies on the Broad Application Test Reactor

International Nuclear Information System (INIS)

Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

1992-08-01

This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

Safety and economic study of special trains

International Nuclear Information System (INIS)

Loscutoff, W.V.; Hall, R.J.

1976-01-01

A comparative evaluation is being conducted of the safety and economics of special (35 mph and less) and regular trains for shipment of spent fuels. The approach, pertinent considerations, and results to date are discussed. The preliminary conclusion is that special train requirements have potential for only a small reduction in the accident likelihood, while increasing the cost

Preliminary safety evaluation for a medical therapy reactor

International Nuclear Information System (INIS)

Jones, J.L.; Neuman, W.A.

1989-01-01

A conceptual design of a passively safe reactor facility for boron neutron capture therapy has been previously described. The medical therapy reactor (MTR) has a maximum power level of 10 MW(thermal) and utilizes 45 wt% uranium in UZrH, 20 wt% 235 U enriched hydride fuel matrix with 1 wt% erbium, which is a burnable poison and provides prompt negative reactivity feedback. The facility has five beam ports for patient treatment and advanced neutron beam research and is capable of 2,000 to 10,000 treatments per year, assuming single 8h/day, 5 day/week operation. The epithermal treatment flux from the beam ports is large, enabling single-session treatment of brain cancers of <10-min duration, with minimal fast neutron and gamma contaminants. The reactor core is designed with sufficient excess reactivity to yield a core lifetime equal to a facility lifetime of 30 yr. A preliminary safety evaluation was performed using the RELAP5 thermal-hydraulic code. The analysis addressed accidents in several major categories, including a pump coastdown, a loss of secondary heat sink, and a $0.5 step reactivity insertion

Posiva's application for a decision in principle concerning a disposal facility for spent nuclear fuel. STUK's statement and preliminary safety appraisal

Energy Technology Data Exchange (ETDEWEB)

Ruokola, E. [ed.

2000-03-01

In May 1999, Posiva Ltd submitted to the Government an application, pursuant to the Nuclear Energy Act, for a Decision in Principle on a disposal facility for spent nuclear fuel from the Finnish nuclear power plants. The Ministry of Trade and Industry requested the Radiation and Nuclear Safety Authority (STUK) to draw up a preliminary safety appraisal concerning the proposed disposal facility. In the beginning of this report, STUK's statement to the Ministry and Industry concerning the proposed disposal facility is given. In that statement, STUK concludes that the Decision in Principle is currently justified from the standpoint of safety. The statement is followed by a safety appraisal, where STUK deems, how the proposed disposal concept, site and facility comply with the safety requirements included in the Government's Decision (478/1999). STUK's preliminary safety appraisal was supported by contributions from a number of outside experts. A collective opinion by an international group of ten distinguished experts is appended to this report. (orig.)

Preliminary safety criteria for organic watch list tanks at the Hanford site

International Nuclear Information System (INIS)

Webb, A.B.; Stewart, J.L.; Turner, O.A.; Plys, M.G.; Malinovic, B.; Grigsby, J.M.; Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J.

1995-11-01

Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended

Preliminary safety criteria for organic watch list tanks at the Hanford site

Energy Technology Data Exchange (ETDEWEB)

Webb, A.B.; Stewart, J.L.; Turner, O.A. [Westinghouse Hanford Co., Richland, WA (United States); Plys, M.G.; Malinovic, B. [Fauske and Associates, Inc., Burr Ridge, IL (United States); Grigsby, J.M. [G & P Consulting, Inc. (United States); Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J. [Pacific Northwest Lab., Portland, OR (United States)

1995-11-01

Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended.

Preliminary Assessment for the Effects of the External Hazard Factors on the Safety of NPPs

International Nuclear Information System (INIS)

Jin, So Beom; Hyun, Seung Gyu; Kim, Sang Yun; Lee, Sung Kyu; Hur, Youl

2010-01-01

The Ch.etsu Offshore Earthquake(2007.7.16) in Japan caused damage to the Kashiwazaki-Kariwa(K-K) Nuclear Power Plants (NPPs) with seismic ground motion that exceeded the design level. This incident drew the interest of the safety evaluation studies for NPPs subjected to earthquakes exceeding the design basis around the world. Also, the Indian Ocean Tsunami(2004.12.26) tripped the Madras NPP by reason of flooding of the intake pump house and inundated the construction site of a fast breeder reactor site in India. In addition, from the various man-made and natural hazards such as the oil spill accident near Mallipo, Taean, Chungnam (2007.12), the forest fire near the Ulchin NPP site, the several inflows of marine organism into the intake of the Ulchin NPP, it was confirmed that the safety of NPPs may be affected by natural and human induced disasters. Intergovernmental Panel on Climate Change (IPCC) has been warned about global warming; the average temperature rose about 1.5 .deg. C during the 20th century and the damages caused by typhoons and heavy rains have also increased in Korea. Accordingly, a natural disaster prevention research team(hereafter team) ,which have been organized and operated since 2009, has assessed the impact of various hazards such as earthquakes and environmental changes due to global warming on the safety of NPP and has discussed to establish countermeasures. This paper introduces that the preliminary assessment for the effects of the external hazard factors on the safety of NPPs was conducted by the team

Research and development of technologies for safe and environmentally optimal recovery and disposal of explosive wastes. Task 2, Preliminary impact assessment for environment, health and safety (EIA)

Energy Technology Data Exchange (ETDEWEB)

Duijm, N.J.; Markert, F. [Risoe (Denmark); Larsen, S.G. [DEMEX A/S (Denmark)

1998-09-01

As described in the project proposal `Research and Development of Technologies for Safe and Environmentally optimal recovery and Disposal of Explosive Wastes`, dated 31. May 1996, the objective of Task 2, Preliminary Impact Assessment for Environment, Health and Safety, is to: Analyse the environmental impact of noise and emissions to air, water and soil; Assess the risk of hazards to workers` health and safety and to the public. Task 2, Preliminary Impact Assessment for Environment, Health and Safety (EIA), has been performed from August 1997 to September 1998. First, a methodology has been established, based on Multi-Criteria Decision Analysis (MCDA), to select the `best` technology on the basis of clearly defined objectives, including minimal impacts on environment, health and safety. This included a review of different types of explosive waste with a focus on the environment implications, identifying the issues relevant to defining the criteria or objectives with respect to environment and safety in the framework of explosive waste, as well as the preliminary definition of objectives for the final impact assessment. Second, the previously identified recovery and disposal technologies (Task 1) have been qualitatively assessed on the basis of the relevant objectives. This qualitative assessment includes also economic considerations and an attempt to rank the technologies in an MCDA framework. (au)

Development of Behavioral Indicators of Competences for Safety Culture of Nuclear Power Plants: A Preliminary Study

International Nuclear Information System (INIS)

Moon, Kwangsu; Kim, Sa Kil; Oh, Yeon Ju; Shin, Youmin; Lee, Yong-Hee; Jang, Tong Il

2015-01-01

The term of safety competency in nuclear field was presented in the OECD/NEA workshop held in 1999. A model of the safety culture competencies in nuclear power plants was developed by KAERI (Korea Atomic Energy Research Institute). In general, a competency (competence) is defined as 'cluster of employee's attribute, knowledge, skill, ability or other characteristic that contributes to successful job performance'. We also defined safety culture competency as 'cluster of various internal characteristics (e.g., knowledge, skill, ability, motive, attitude and etc.) of employee that contribute to perform job safely and shape a healthy and strong safety culture.' By this definition, the safety culture competency is the broader construct including job competency. An employee having high level of safety culture competency shows extra discretionary effort to improve safety of peer, team and organization in addition to the individual's successful and safe job accomplishment. The behavioral indicators for each of the competencies are focal points of conversations on progress and are monitored continuously by self-assessment and managers or supervisors' intervention. Deficiencies in any of these indicators can point to coaching, training or other learning opportunities that employees may be required in order to improve. The purpose of this study was to derive a model of safety competencies for improving safety culture of NPPs and develop a set of behavioral indicators of each competency. In addition, the method of measuring behavioral indicators was suggested. For the application of developed safety culture competences and behavioral indicators, the most suitable measuring method for behavioral indicators must be developed. In the case of behavioral observations, behavioral dimensions (frequency, persistence and latency), observation possibility, occurrence basis of behavior (daily job performance, situational dependent) are considered to

Development of Behavioral Indicators of Competences for Safety Culture of Nuclear Power Plants: A Preliminary Study

Energy Technology Data Exchange (ETDEWEB)

Moon, Kwangsu; Kim, Sa Kil; Oh, Yeon Ju; Shin, Youmin; Lee, Yong-Hee; Jang, Tong Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The term of safety competency in nuclear field was presented in the OECD/NEA workshop held in 1999. A model of the safety culture competencies in nuclear power plants was developed by KAERI (Korea Atomic Energy Research Institute). In general, a competency (competence) is defined as 'cluster of employee's attribute, knowledge, skill, ability or other characteristic that contributes to successful job performance'. We also defined safety culture competency as 'cluster of various internal characteristics (e.g., knowledge, skill, ability, motive, attitude and etc.) of employee that contribute to perform job safely and shape a healthy and strong safety culture.' By this definition, the safety culture competency is the broader construct including job competency. An employee having high level of safety culture competency shows extra discretionary effort to improve safety of peer, team and organization in addition to the individual's successful and safe job accomplishment. The behavioral indicators for each of the competencies are focal points of conversations on progress and are monitored continuously by self-assessment and managers or supervisors' intervention. Deficiencies in any of these indicators can point to coaching, training or other learning opportunities that employees may be required in order to improve. The purpose of this study was to derive a model of safety competencies for improving safety culture of NPPs and develop a set of behavioral indicators of each competency. In addition, the method of measuring behavioral indicators was suggested. For the application of developed safety culture competences and behavioral indicators, the most suitable measuring method for behavioral indicators must be developed. In the case of behavioral observations, behavioral dimensions (frequency, persistence and latency), observation possibility, occurrence basis of behavior (daily job performance, situational dependent) are considered to

Preliminary systems-interaction results from the Digraph Matrix Analysis of the Watts Bar Nuclear Power Plant safety-injection systems

International Nuclear Information System (INIS)

Sacks, I.J.; Ashmore, B.C.; Champney, J.M.; Alesso, H.P.

1983-06-01

This report provides preliminary results generated by a Digraph Matrix Analysis (DMA) for a Systems Interaction analysis performed on the Safety Injection System of the Tennessee Valley Authority Watts Bar Nuclear Power Plant. An overview of DMA is provided along with a brief description of the computer codes used in DMA

Safety report content and development for test loop facility on MARIA reactor

International Nuclear Information System (INIS)

Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

1982-01-01

A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

Preliminary safety evaluation for the plutonium stabilization and packaging system

International Nuclear Information System (INIS)

Shapley, J.E.

1997-01-01

This Preliminary Safety Evaluation (PSE) describes and analyzes the installation and operation of the Plutonium Stabilization and Packaging System (SPS) at the Plutonium Finishing Plant (PFP). The SPS is a combination of components required to expedite the safe and timely storage of Plutonium (Pu) oxide. The SPS program will receive site Pu packages, process the Pu for storage, package the Pu into metallic containers, and safely store the containers in a specially modified storage vault. The location of the SPS will be in the 2736- ZB building and the storage vaults will be in the 2736-Z building of the PFP, as shown in Figure 1-1. The SPS will produce storage canisters that are larger than those currently used for Pu storage at the PFP. Therefore, the existing storage areas within the PFP secure vaults will require modification. Other modifications will be performed on the 2736-ZB building complex to facilitate the installation and operation of the SPS

1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

International Nuclear Information System (INIS)

Blomeke, J.O.

1977-08-01

This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations

1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

Energy Technology Data Exchange (ETDEWEB)

Blomeke, J.O.

1977-08-01

This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations. (DLC)

Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Timothy Solack; Carol Mason

2012-03-01

A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

Preliminary assessments the shortcut to remediation (category III-surplus facility assessments)

International Nuclear Information System (INIS)

Byars, L.L.

1995-01-01

This report presents the details of the preliminary assessments for the shortcut of decontamination of surplus nuclear facilities. Topics discussed include: environment, health and safety concerns; economic considerations; reduction of transition time; preliminary characterization reports; preliminary project plan; health and safety plan; quality assurance plan; surveillance and maintenance plan; and waste management plan

Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

Energy Technology Data Exchange (ETDEWEB)

Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-10-01

Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.

Non-clinical studies in the process of new drug development - Part II: Good laboratory practice, metabolism, pharmacokinetics, safety and dose translation to clinical studies.

Science.gov (United States)

Andrade, E L; Bento, A F; Cavalli, J; Oliveira, S K; Schwanke, R C; Siqueira, J M; Freitas, C S; Marcon, R; Calixto, J B

2016-12-12

The process of drug development involves non-clinical and clinical studies. Non-clinical studies are conducted using different protocols including animal studies, which mostly follow the Good Laboratory Practice (GLP) regulations. During the early pre-clinical development process, also known as Go/No-Go decision, a drug candidate needs to pass through several steps, such as determination of drug availability (studies on pharmacokinetics), absorption, distribution, metabolism and elimination (ADME) and preliminary studies that aim to investigate the candidate safety including genotoxicity, mutagenicity, safety pharmacology and general toxicology. These preliminary studies generally do not need to comply with GLP regulations. These studies aim at investigating the drug safety to obtain the first information about its tolerability in different systems that are relevant for further decisions. There are, however, other studies that should be performed according to GLP standards and are mandatory for the safe exposure to humans, such as repeated dose toxicity, genotoxicity and safety pharmacology. These studies must be conducted before the Investigational New Drug (IND) application. The package of non-clinical studies should cover all information needed for the safe transposition of drugs from animals to humans, generally based on the non-observed adverse effect level (NOAEL) obtained from general toxicity studies. After IND approval, other GLP experiments for the evaluation of chronic toxicity, reproductive and developmental toxicity, carcinogenicity and genotoxicity, are carried out during the clinical phase of development. However, the necessity of performing such studies depends on the new drug clinical application purpose.

Some safety studies for conceptual fusion--fission hybrid reactors. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.

1978-07-01

The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

Preliminary conceptual study of engineering-scale pyroprocess demonstration facility

International Nuclear Information System (INIS)

Moon, Seong-In; Chong, Won-Myung; You, Gil-Sung; Ku, Jeong-Hoe; Kim, Ho-Dong

2013-01-01

Highlights: ► The conceptual design of a pyroprocess demonstration facility was performed. ► The design requirements for the pyroprocess hot cell and equipment were determined. ► The maintenance concept for the pyroprocess hot cell was presented. -- Abstract: The development of an effective management technology of spent fuel is important to enhance environmental friendliness, cost viability and proliferation resistance. In Korea, pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems. PRIDE (PyRoprocess Integrated inactive DEmonstration facility) has been developed from 2007 to 2012 in Korea as a cold test facility to support integrated pyroprocessing and an equipment demonstration, which is essential to verify the pyroprocess technology. As the next stage of PRIDE, the design requirements of an engineering-scale demonstration facility are being developed, and the preliminary conceptual design of the facility is being performed for the future. In this paper, the main design requirements for the engineering-scale pyroprocess demonstration facility were studied in the throughput of 10tHM a year. For the preliminary conceptual design of the facility, the design basis of the pyroprocess hot cell was suggested, and the main equipment, main process area, operation area, maintenance area, and so on were arranged in consideration of the effective operation of the hot cells. Also, the argon system was designed to provide and maintain a proper inert environment for the pyroprocess. The preliminary conceptual design data will be used to review the validity of the engineering-scale pyroprocess demonstration facility that enhances both safety and nonproliferation

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2015-01-30

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

White paper: Preliminary assessment of LNG vehicle technology, economics, and safety issues (Revision 1). Topical report, April-August 1991

International Nuclear Information System (INIS)

Powars, C.; Lucher, D.; Moyer, C.; Browning, L.

1992-01-01

The objective of the study is to evaluate the potential of LNG as a vehicle fuel, to determine market niches, and to identify needed technology improvements. The white paper is being issued when the work is approximately 30 percent complete to preview the study direction, draw preliminary conclusions, and make initial recommendations. Interim findings relative to LNG vehicle technology, economics, and safety are presented. It is important to decide if heavier hydrocarbons should be allowed in LNG vehicle fuel. Development of suitable refueling couplings and vehicle fuel supply pressure systems are recommended. Initial economics analyses considered transit buses and pickup and delivery trucks fueled via onsite liquefiers and imported LNG. Net user costs were more than (but in some cases close to) those for diesel fuel and gasoline. Lowering the cost of small-scale liquefiers would significantly improve the economics of LNG vehicles. New emissions regulations may introduce considerations beyond simple cost comparisons. LNG vehicle safety and available accident data are reviewed. Consistent codes for LNG vehicles and refueling facilities are needed

Early Childhood Safety Education: An Overview of Safety Curriculum in Outer Metropolitan, Regional and Rural NSW

Science.gov (United States)

Barr, Jennifer; Saltmarsh, Sue; Klopper, Christopher

2009-01-01

This article reports on preliminary findings from a 2008 survey and telephone interviews with 27 directors of early childhood education and care (ECEC) services located in regional and rural districts of the Australian state of New South Wales. Data from the study suggests that some areas of safety education--most notably road/traffic safety and…

Raised crosswalks on entrance to the roundabout-a case study on effectiveness of treatment on pedestrian safety and convenience.

Science.gov (United States)

Candappa, Nimmi; Stephan, Karen; Fotheringham, Nicola; Lenné, Michael G; Corben, Bruce

2014-01-01

A common concern in the use of a roundabout is providing adequately for the pedestrian. This unique roundabout layout, which introduces raised crosswalks directly at the roundabout entrance, as opposed to at a car length back, aims at improving safety and convenience for pedestrians at roundabouts. A preliminary evaluation of the layout was undertaken to establish its effectiveness in meeting study objectives. A quasi-experimental before-and-after study design was used to compare speeds on approach and immediately prior to the crossing to ascertain potential impact speed and implications for pedestrian safety. Compliance to crossing and crossing time were also compared in relation to safety and convenience outcomes. A questionnaire assessed pedestrian perception of the safety and convenience at the roundabout before and after treatment. Results from this case study indicate that mean approach speeds (free speeds 30 m from crossing) reduced from 32.7 to 30.7 km/h and immediately prior to crossing, mean speeds reduced from 19.1 to 16.3 km/h. There was also a marked reduction in proportions of vehicles traveling at speeds that could elevate risk to pedestrians. Total crossing time after treatment reduced by around 4 s, and crossing compliance increased from approximately half to approximately 90 percent. Survey of pedestrians indicated positive response to the perceived safety and convenience posttreatment. Preliminary results of the case study suggest positive safety and convenience outcomes. Implications for pedestrian safety include less exposure to traffic and lower risk of serious injury, particularly for elderly pedestrians; convenience outcomes include shorter waiting times to cross and greater compliance to the crossing. A larger study is required to substantiate the findings.

Deep Borehole Disposal Safety Analysis.

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2016-10-01

This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

Safety barriers and safety functions a comparison of different applications

International Nuclear Information System (INIS)

Harms-Ringdahl, L.

1998-01-01

A study is being made with the focus on different theories and applications concerning 'safety barriers' and 'safety functions'. One aim is to compare the characteristics of different kinds of safely functions, which can be purpose, efficiency, reliability, weak points etc. A further aim is to summarize how the combination of different barriers are described and evaluated. Of special interest are applications from nuclear and chemical process safety. The study is based on a literature review, interviews and discussions. Some preliminary conclusions are made. For example, it appears to exist a need for better tools to support the design and evaluation of procedures. There are a great number of theoretical models describing safety functions. However, it still appears to be an interest in further development of models, which might give the basis for improved practical tools. (author)

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)

French 900 MWe PWR PSA preliminary results

International Nuclear Information System (INIS)

Lanore, J.M.; Brisbois, J.

1988-10-01

A PSA is performed by the Safety Assessment Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the relative preliminary results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures

Preliminary Safety Information Document for the Standard MHTGR. Volume 1, (includes latest Amendments)

Energy Technology Data Exchange (ETDEWEB)

None

1986-01-01

With NRC concurrence, the Licensing Plan for the Standard HTGR describes an application program consistent with 10CFR50, Appendix O to support a US Nuclear Regulatory Commission (NRC) review and design certification of an advanced Standard modular High Temperature Gas-Cooled Reactor (MHTGR) design. Consistent with the NRC's Advanced Reactor Policy, the Plan also outlines a series of preapplication activities which have as an objective the early issuance of an NRC Licensability Statement on the Standard MHTGR conceptual design. This Preliminary Safety Information Document (PSID) has been prepared as one of the submittals to the NRC by the US Department of Energy in support of preapplication activities on the Standard MHTGR. Other submittals to be provided include a Probabilistic Risk Assessment, a Regulatory Technology Development Plan, and an Emergency Planning Bases Report.

Preliminary report in radiological safety for 1993 hydrology campaign

International Nuclear Information System (INIS)

Badano, A.; Suraez, R.; Dellepere, A.; Barreiro, M.

1993-01-01

The purpose of this report is to provide a study about industrial effluents influence on water pollution of Montevideo coastal beaches. The methods which have been considered are nuclear tracer techniques with a special attention in the radioprotection supervision. Three points are considered as evaluation: handling of radioactive tracers and safety, radiation protection workers, environment and public safety. tabs

Preliminary Integrated Safety Analysis Status Report

International Nuclear Information System (INIS)

Gwyn, D.

2001-01-01

This report provides the status of the potential Monitored Geologic Repository (MGR) Integrated Safety Analysis (EA) by identifying the initial work scope scheduled for completion during the ISA development period, the schedules associated with the tasks identified, safety analysis issues encountered, and a summary of accomplishments during the reporting period. This status covers the period from October 1, 2000 through March 30, 2001

Preliminary Slope Stability Study Using Slope/ W

International Nuclear Information System (INIS)

Nazran Harun; Mohd Abd Wahab Yusof; Kamarudin Samuding; Mohd Muzamil Mohd Hashim; Nurul Fairuz Diyana Bahrudin

2014-01-01

Analyzing the stability of earth structures is the oldest type of numerical analysis in geotechnical engineering. Limit equilibrium types of analyses for assessing the stability of earth slopes have been in use in geotechnical engineering for many decades. Modern limit equilibrium software is making it possible to handle ever-increasing complexity within an analysis. It is being considered as the potential method in dealing with complex stratigraphy, highly irregular pore-water pressure conditions, various linear and nonlinear shear strength models and almost any kind of slip surface shape. It allows rapid decision making by providing an early indication of the potential suitability of sites based on slope stability analysis. Hence, a preliminary slope stability study has been developed to improve the capacity of Malaysian Nuclear Agency (Nuclear Malaysia) in assessing potential sites for Borehole Disposal for Disused Sealed Radioactive Sources. The results showed that geometry of cross section A-A ' , B-B ' , C-C ' and D-D ' achieved the factor of safety not less than 1.4 and these are deemed acceptable. (author)

Preliminary study of the safety and efficacy of medium-chain triglycerides for use as an intraocular tamponading agent in minipigs.

Science.gov (United States)

Soler, Vincent J; Laurent, Camille; Sakr, Frédéric; Regnier, Alain; Tricoire, Cyrielle; Cases, Olivier; Kozyraki, Renata; Douet, Jean-Yves; Pagot-Mathis, Véronique

2017-08-01

To date, only silicone oils and gases have the appropriate characteristics for use in vitreo-retinal surgery as vitreous substitutes with intraocular tamponading properties. This preliminary study evaluated the safety and efficacy of medium-chain triglycerides (MCTs) for use as a tamponading agent in minipigs. In 15 minipigs, 15 right eyes underwent vitrectomies followed by injection of MCT tamponade (day 1). Two groups were defined. In Group A (ten eyes), the surgical procedure before MCT injection included induced rhegmatogenous retinal detachment (RRD), retina flattening, and retinopexy. In Group B (five eyes), MCT was injected without inducing RRD; in these eyes, MCT was removed on day 90. Pigs were sacrificed on day 45 (Group A) or 120 (Group B). Eyes were examined on days 1, 5, 15, and 45 in both groups and on days 90 and 120 in Group B. In Group B only, we performed bilateral electroretinography examinations on days 1 and 120, and histological examinations of MCTs and controlateral eyes were performed after sacrifice. In Group A eyes (n = 9; one eye was non-assessable), on day 45, the retina was flat in seven eyes and two RRD detachments were observed in insufficiently MCT-filled eyes. In Group B, electroretinography showed no significant differences between MCT eyes and controls on days 1 or 120. Histological analyses revealed no signs of retinal toxicity. This study showed that MCT tamponade seems to be effective and safe; however, additional studies are needed before it becomes commonly used as a tamponading agent in humans.

Liquid metal reactor development -Studies on safety measure of LMR coolant

International Nuclear Information System (INIS)

Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung

1995-07-01

A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author)

Liquid metal reactor development -Studies on safety measure of LMR coolant

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

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)

A preliminary study on the design in architecture of nuclear and radiation safety standard system

International Nuclear Information System (INIS)

Song Dahu; Zhang Chi; Yang Lili; Li Bin; Liu Yingwei; An Hongzhen; Gao Siyi; Liu Ting; Meng De

2014-01-01

The connotation and function of nuclear and radiation safety standards are analyzed, and their relationships with the relevant laws and regulations are discussed in the paper. Some suggestions and blue print of overall architecture to build nuclear and radiation safety standard system are proposed, on the basis of researching the application status quo, existing problems and needs for nuclear and radiation safety standards in China. This work is a beneficial exploration and attempt to establish China's nuclear and radiation safety standards. (authors)

Preliminary experimental study of liquid lithium water interaction

International Nuclear Information System (INIS)

You, X.M.; Tong, L.L.; Cao, X.W.

2015-01-01

Highlights: • Explosive reaction occurs when lithium temperature is over 300 °C. • The violence of liquid lithium water interaction increases with the initial temperature of liquid lithium. • The interaction is suppressed when the initial water temperature is above 70 °C. • Steam explosion is not ignorable in the risk assessment of liquid lithium water interaction. • Explosion strength of liquid lithium water interaction is evaluated by explosive yield. - Abstract: Liquid lithium is the best candidate for a material with low Z and low activation, and is one of the important choices for plasma facing materials in magnetic fusion devices. However, liquid lithium reacts violently with water under the conditions of loss of coolant accidents. The release of large heats and hydrogen could result in the dramatic increase of temperature and pressure. The lithium–water explosion has large effect on the safety of fusion devices, which is an important content for the safety assessment of fusion devices. As a preliminary investigation of liquid lithium water interaction, the test facility has been built and experiments have been conducted under different conditions. The initial temperature of lithium droplet ranged from 200 °C to 600 °C and water temperature was varied between 20 °C and 90 °C. Lithium droplets were released into the test section with excess water. The shape of lithium droplet and steam generated around the lithium were observed by the high speed camera. At the same time, the pressure and temperature in the test section were recorded during the violent interactions. The preliminary experimental results indicate that the initial temperature of lithium and water has an effect on the violence of liquid lithium water interaction.

Preliminary experimental study of liquid lithium water interaction

Energy Technology Data Exchange (ETDEWEB)

You, X.M.; Tong, L.L.; Cao, X.W., E-mail: caoxuewu@sjtu.edu.cn

2015-10-15

Highlights: • Explosive reaction occurs when lithium temperature is over 300 °C. • The violence of liquid lithium water interaction increases with the initial temperature of liquid lithium. • The interaction is suppressed when the initial water temperature is above 70 °C. • Steam explosion is not ignorable in the risk assessment of liquid lithium water interaction. • Explosion strength of liquid lithium water interaction is evaluated by explosive yield. - Abstract: Liquid lithium is the best candidate for a material with low Z and low activation, and is one of the important choices for plasma facing materials in magnetic fusion devices. However, liquid lithium reacts violently with water under the conditions of loss of coolant accidents. The release of large heats and hydrogen could result in the dramatic increase of temperature and pressure. The lithium–water explosion has large effect on the safety of fusion devices, which is an important content for the safety assessment of fusion devices. As a preliminary investigation of liquid lithium water interaction, the test facility has been built and experiments have been conducted under different conditions. The initial temperature of lithium droplet ranged from 200 °C to 600 °C and water temperature was varied between 20 °C and 90 °C. Lithium droplets were released into the test section with excess water. The shape of lithium droplet and steam generated around the lithium were observed by the high speed camera. At the same time, the pressure and temperature in the test section were recorded during the violent interactions. The preliminary experimental results indicate that the initial temperature of lithium and water has an effect on the violence of liquid lithium water interaction.

Preliminary Study for Application of the New Safety Goal related with the Limitation of Cs-137 release

Energy Technology Data Exchange (ETDEWEB)

Seo, Mi Ro; Shin, Tae Young [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In the New Safety Goal, it is clearly stated that the Probabilistic Safety Assessment (PSA) should be performed with the proper technical appropriateness, the detail, and the scope in accordance with the integrated risk assessment against the accident for the nuclear power plants. This requirement is known to be come from the provision for preventing the long term ground contamination due to the release of radioactive material. However, there were so many concerns that this goal is so severe that the current design, even in the case of the constructing nuclear power plants, cannot meet this criterion. Especially for the operating nuclear power plants, since there were no mitigation facilities against the severe accident at the design stages, the application of this new goal is known to be much severe that the constructing nuclear power plants and it is necessary to develop the alternative methods to strengthen the safety of the operating nuclear power plants. The purpose of this study is to review the new safety goal from the view point of severe accident analysis and probabilistic safety assessment, and to find the appropriate methods in order to meet that goal for the operating nuclear power plants. In order to strengthen the safety for domestic nuclear power plants, all of the domestic operating nuclear power plants are required to prepare the Accident Management Plan within 3 years. Also, this Accident Management Plan should meet the New Safety Goal including the requirement that the sum of the accident frequency that the release of the radioactive nuclide Cs- 137 to the environment exceeds the 100TBq should be less than 1.0E-6/RY. Since the operating nuclear power plants was not designed against the severe accident and they have the limited exclusive mitigation facilities, it is not easy to meet the New Safety Goal. So, it is necessary to develop the alternative methods to meet the New Safety Goal. In this study, the amount of Cs-137 released to the

Preliminary Study for Application of the New Safety Goal related with the Limitation of Cs-137 release

International Nuclear Information System (INIS)

Seo, Mi Ro; Shin, Tae Young

2016-01-01

In the New Safety Goal, it is clearly stated that the Probabilistic Safety Assessment (PSA) should be performed with the proper technical appropriateness, the detail, and the scope in accordance with the integrated risk assessment against the accident for the nuclear power plants. This requirement is known to be come from the provision for preventing the long term ground contamination due to the release of radioactive material. However, there were so many concerns that this goal is so severe that the current design, even in the case of the constructing nuclear power plants, cannot meet this criterion. Especially for the operating nuclear power plants, since there were no mitigation facilities against the severe accident at the design stages, the application of this new goal is known to be much severe that the constructing nuclear power plants and it is necessary to develop the alternative methods to strengthen the safety of the operating nuclear power plants. The purpose of this study is to review the new safety goal from the view point of severe accident analysis and probabilistic safety assessment, and to find the appropriate methods in order to meet that goal for the operating nuclear power plants. In order to strengthen the safety for domestic nuclear power plants, all of the domestic operating nuclear power plants are required to prepare the Accident Management Plan within 3 years. Also, this Accident Management Plan should meet the New Safety Goal including the requirement that the sum of the accident frequency that the release of the radioactive nuclide Cs- 137 to the environment exceeds the 100TBq should be less than 1.0E-6/RY. Since the operating nuclear power plants was not designed against the severe accident and they have the limited exclusive mitigation facilities, it is not easy to meet the New Safety Goal. So, it is necessary to develop the alternative methods to meet the New Safety Goal. In this study, the amount of Cs-137 released to the

Safety and preliminary evidence of biologic efficacy of a mammaglobin-a DNA vaccine in patients with stable metastatic breast cancer.

Science.gov (United States)

Tiriveedhi, Venkataswarup; Tucker, Natalia; Herndon, John; Li, Lijin; Sturmoski, Mark; Ellis, Matthew; Ma, Cynthia; Naughton, Michael; Lockhart, A Craig; Gao, Feng; Fleming, Timothy; Goedegebuure, Peter; Mohanakumar, Thalachallour; Gillanders, William E

2014-12-01

Mammaglobin-A (MAM-A) is overexpressed in 40% to 80% of primary breast cancers. We initiated a phase I clinical trial of a MAM-A DNA vaccine to evaluate its safety and biologic efficacy. Patients with breast cancer with stable metastatic disease were eligible for enrollment. Safety was monitored with clinical and laboratory assessments. The CD8 T-cell response was measured by ELISPOT, flow cytometry, and cytotoxicity assays. Progression-free survival (PFS) was described using the Kaplan-Meier product limit estimator. Fourteen subjects have been treated with the MAM-A DNA vaccine and no significant adverse events have been observed. Eight of 14 subjects were HLA-A2(+), and the CD8 T-cell response to vaccination was studied in detail. Flow cytometry demonstrated a significant increase in the frequency of MAM-A-specific CD8 T cells after vaccination (0.9% ± 0.5% vs. 3.8% ± 1.2%; P cells (41 ± 32 vs. 215 ± 67 spm; P cell responses, and preliminary evidence suggests improved PFS. Additional studies are required to define the potential of the MAM-A DNA vaccine for breast cancer prevention and/or therapy. ©2014 American Association for Cancer Research.

Socio-technological study for establishing comprehensive nuclear safety system

International Nuclear Information System (INIS)

Furuta, Kazuo; Kanno, Taro; Yagi, Ekou; Shuto, Yuki

2003-01-01

This paper presents an overview and preliminary results of a research project on social-technology for nuclear safety, which started in October 2001. In particular, emergency response preparedness against nuclear disaster and consensus development will be discussed. The architecture of an emergency response simulator will be given, which is for assessing design of disaster prevention systems. A conceptual model of evacuation behavior of a resident has been constructed from analysis of past disaster cases. As for consensus development, deliberation spaces of actual committee meetings were constructed by analyzing transcripts of the meetings based on an opinion schema. A model of consensus development process has been proposed from the traces of participants' opinions over the deliberation spaces. Such a socio-technological approach will be useful not only for nuclear safety but also for safety of non-nuclear domains and human activities of a high hazard potential; it is expected to contribute to establishing risk-aware society of the future. (author)

Preliminary study of impact fragility to RC wall subjected to aircraft impact

International Nuclear Information System (INIS)

Shin, Sang Shup; Hahm, Dae Gi; Choi, In Kil

2012-01-01

International experience has shown that internal and external hazards such as fires, earthquakes, and aircraft impacts can be significant safety contributors to the risk to infrastructures such as nuclear power plants. Since the aircraft accident at the World Trade Center (WTC) on September 11, 2001, an aircraft impact problem has been increasingly of the interest and is one of important categories of an unexpected external hazard field. To date, aircraft impact analyses has most focused on the response analysis to the target structures. However, this preliminary study carried out an impact fragility analysis to reinforced concrete (RC) wall subjected to an aircraft impact. The aircraft velocity is used as the important variable of this study. The impact analysis of the applied Ri era's forcing function is used by Abaqus/Explicit

Preliminary risk assessments of the small HTGR

International Nuclear Information System (INIS)

Everline, C.J.; Bellis, E.A.

1985-05-01

Preliminary investment and safety risk assessments were performed for a preconceptual design of a four-module 250-MW(t) side-by-side steel-vessel pebble bed HTGR plant. Broad event spectra were analyzed involving component damage resulting in unscheduled plant outages and fission product releases resulting in offsite doses. The preliminary assessment indicates at this stage of the design that two categories of events govern the investment risk envelope: primary coolant leaks which release some circulating and plate-out activity that contaminates the confinement and turbogenerator damage which involves extensive turbine blade failure. Primary coolant leaks are important contributors because associated cleanup and decontamination requirements result in longer outages that arise from other events with comparable frequencies. Turbogenerator damage is the salient low-frequency investment risk accident due to the relatively long outages being experienced in the industry. Thermal transients are unimportant investment risk contributors because pressurized core heatups cause little damage, and depressurized core heatups occur at negligible frequencies relative to the forced outage goal. These preliminary results demonstrate investment and safety risk goal compliance at this stage in the design process. Studies are continuing in order to provide valuable insights into risk-significant events to assure a balanced approach to meeting user and regulatory requirements

A study on a reliability assessment methodology for the VHTR safety systems

International Nuclear Information System (INIS)

Lee, Hyung Sok

2012-02-01

The passive safety system of a 300MWt VHTR (Very High Temperature Reactor)which has attracted worldwide attention recently is actively considered for designing the improvement in the safety of the next generation nuclear power plant. The passive system functionality does not rely on an external source of the electrical support system,but on an intelligent use of the natural phenomena, such as convection, conduction, radiation, and gravity. It is not easy to evaluate quantitatively the reliability of the passive safety for the risk analysis considering the existing active system failure since the classical reliability assessment method could not be applicable. Therefore a new reliability methodology needs to be developed and applied for evaluating the reliability of the conceptual designed VHTR in this study. The preliminary evaluation and conceptualization are performed using the concept of the load and capacity theory related to the reliability physics model. The method of response surface method (RSM) is also utilized for evaluating the maximum temperature of nuclear fuel in this study. The significant variables and their correlation are considered for utilizing the GAMMA+ code. The proposed method might contribute to designing the new passive system of the VHTR

Job safety and awareness analysis of safety implementation among electrical workers in airport service company

Directory of Open Access Journals (Sweden)

Putra Perdana Suteja

2018-01-01

Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

On the safety of aircraft systems: A case study

Energy Technology Data Exchange (ETDEWEB)

Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

1997-05-14

An airplane is a highly engineered system incorporating control- and feedback-loops which often, and realistically, are non-linear because the equations describing such feedback contain products of state variables, trigonometric or square-root functions, or other types of non-linear terms. The feedback provided by the pilot (crew) of the airplane also is typically non-linear because it has the same mathematical characteristics. An airplane is designed with systems to prevent and mitigate undesired events. If an undesired triggering event occurs, an accident may process in different ways depending on the effectiveness of such systems. In addition, the progression of some accidents requires that the operating crew take corrective action(s), which may modify the configuration of some systems. The safety assessment of an aircraft system typically is carried out using ARP (Aerospace Recommended Practice) 4761 (SAE, 1995) methods, such as Fault Tree Analysis (FTA) and Failure Mode and Effects Analysis (FMEA). Such methods may be called static because they model an aircraft system on its nominal configuration during a mission time, but they do not incorporate the action(s) taken by the operating crew, nor the dynamic behavior (non-linearities) of the system (airplane) as a function of time. Probabilistic Safety Assessment (PSA), also known as Probabilistic Risk Assessment (PRA), has been applied to highly engineered systems, such as aircraft and nuclear power plants. PSA encompasses a wide variety of methods, including event tree analysis (ETA), FTA, and common-cause analysis, among others. PSA should not be confused with ARP 4761`s proposed PSSA (Preliminary System Safety Assessment); as its name implies, PSSA is a preliminary assessment at the system level consisting of FTA and FMEA.

Safety study application guide

International Nuclear Information System (INIS)

1993-07-01

Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

EXPLOSION POTENTIAL ASSESSMENT OF HEAT EXCHANGER NETWORK AT THE PRELIMINARY DESIGN STAGE

Directory of Open Access Journals (Sweden)

MOHSIN PASHA

2016-07-01

Full Text Available The failure of Shell and Tube Heat Exchangers (STHE is being extensively observed in the chemical process industries. This failure can cause enormous production loss and have a potential of dangerous consequences such as an explosion, fire and toxic release scenarios. There is an urgent need for assessing the explosion potential of shell and tube heat exchanger at the preliminary design stage. In current work, inherent safety index based approach is used to resolve the highlighted issue. Inherent Safety Index for Shell and Tube Heat Exchanger (ISISTHE is a newly developed index for assessing the inherent safety level of a STHE at the preliminary design stage. This index is composed of preliminary design variables and integrated with the process design simulator (Aspen HYSYS. Process information can easily be transferred from process design simulator to MS Excel spreadsheet owing to this integration. This index could potentially facilitate the design engineer to analyse the worst heat exchanger in the heat exchanger network. Typical heat exchanger network of the steam reforming process is presented as a case study and the worst heat exchanger of this network has been identified. It is inferred from this analysis that shell and tube heat exchangers possess high operating pressure, corrected mean temperature difference (CMTD and flammability and reactive potential needs to be critically analysed at the preliminary design stage.

The association between EMS workplace safety culture and safety outcomes.

Science.gov (United States)

Weaver, Matthew D; Wang, Henry E; Fairbanks, Rollin J; Patterson, Daniel

2012-01-01

Prior studies have highlighted wide variation in emergency medical services (EMS) workplace safety culture across agencies. To determine the association between EMS workplace safety culture scores and patient or provider safety outcomes. We administered a cross-sectional survey to EMS workers affiliated with a convenience sample of agencies. We recruited these agencies from a national EMS management organization. We used the EMS Safety Attitudes Questionnaire (EMS-SAQ) to measure workplace safety culture and the EMS Safety Inventory (EMS-SI), a tool developed to capture self-reported safety outcomes from EMS workers. The EMS-SAQ provides reliable and valid measures of six domains: safety climate, teamwork climate, perceptions of management, working conditions, stress recognition, and job satisfaction. A panel of medical directors, emergency medical technicians and paramedics, and occupational epidemiologists developed the EMS-SI to measure self-reported injury, medical errors and adverse events, and safety-compromising behaviors. We used hierarchical linear models to evaluate the association between EMS-SAQ scores and EMS-SI safety outcome measures. Sixteen percent of all respondents reported experiencing an injury in the past three months, four of every 10 respondents reported an error or adverse event (AE), and 89% reported safety-compromising behaviors. Respondents reporting injury scored lower on five of the six domains of safety culture. Respondents reporting an error or AE scored lower for four of the six domains, while respondents reporting safety-compromising behavior had lower safety culture scores for five of the six domains. Individual EMS worker perceptions of workplace safety culture are associated with composite measures of patient and provider safety outcomes. This study is preliminary evidence of the association between safety culture and patient or provider safety outcomes.

R and D Requirements, RF Gun Mode Studies, FEL-2 Steady-State Studies, Preliminary FEL-1 Time-Dependent Studies, and Preliminary Layout Option Investigation

International Nuclear Information System (INIS)

Byrd, John; Corlett, John; Doolittle, Larry; Fawley, William; Lidia, Steven; Penn, Gregory; Ratti, Alex; Staples, John; Wilcox Russell; Wurtele, Jonathan; Zholents, Alexander

2005-01-01

This report constitutes the third deliverable of LBNLs contracted role in the FERMI (at) Elettra Technical Optimization study. It describes proposed RandD activities for the baseline design of the Technical Optimization Study, initial studies of the RF gun mode-coupling and potential effects on beam dynamics, steady-state studies of FEL-2 performance to 10 nm, preliminary studies of time-dependent FEL-1 performance using electron bunch distribution from the start-to-end studies, and a preliminary investigation of a configuration with FEL sinclined at a small angle from the line of the linac

Synthesis of the safety studies carried out on the GFR2400

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108, Saint Paul-lez-Durance (France); Bassi, C. [CEA, DEN, DER, F-13108, Saint Paul-lez-Durance (France); Bentivoglio, F. [CEA, DEN, DM2S, F-38054, Grenoble (France); Audubert, F. [CEA, DEN, DEC, F-13108, Saint Paul-lez-Durance (France); Gueneau, C. [CEA, DEN, DPC, F-91191, Gif-sur-yvette (France); Rimpault, G. [CEA, DEN, DER, F-13108, Saint Paul-lez-Durance (France); Journeau, C. [CEA, DEN, DTN, F-13108, Saint Paul-lez-Durance (France)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Insights from accident studies and PSA have consolidated GFR2400 design. Black-Right-Pointing-Pointer Safety margins are adequate for design basis accidents. Black-Right-Pointing-Pointer Core cooling strategy is reinforced by use of PCS for frequent events. Black-Right-Pointing-Pointer Prevention of core degradation is shown in challenging hypothetic situations. Black-Right-Pointing-Pointer It is shown that most of severe accidents can be managed despite limited test data. - Abstract: The present paper is dedicated to the synthesis of the safety studies carried out on the 2400 MWth gas-cooled fast reactor (GFR2400) concept developed at CEA. The analysis of the reference design basis accidents investigated up to now, has shown margins up to the acceptance criteria, equal at least to 300 Degree-Sign C for the category 3 situations and larger than 100 Degree-Sign C for the category 4 situations. The dimensioning of the decay heat removal (DHR) loops and of the power conversion system (PCS) loops has been shown adequate even for bounding degraded situations including multiple failures. Furthermore, in the following part of the paper, it is shown how the main insights provided by a level 1 probabilistic safety assessment (PSA) carried out at an early stage of the design, have led to reinforce the reliability of the DHR function in high pressure conditions by using the PCS as the first mean to cool the core; in the same time, on the basis of a combination of deterministic augments and of PSA results, a design simplification process has led to add a low pressure DHR loop to replace a high pressure DHR loop. The last section is dedicated to prevention and preliminary study of severe accidents (SA). Four SA families have been identified depending on the dynamics and on the scale of the considered accident. The possibility to prevent core degradation by using an adapted accident management (nitrogen injection, use of PCS loops) has

Preliminary report of radiological safety to hydrology 1993 campaign

International Nuclear Information System (INIS)

Badano, A.; Suarez Antola, R.; Dellepere, A.; Barreiro, M.

1993-01-01

This report has been prepared based on the interaction between project managers and division radiological Protection and Nuclear Safety. In seeking to establish a basis for approval from the point of view of radiation safety practices . The idea for the audit has been provided at all times because the interest was the exchange of ideas and the use of common sense to improve the safety of radioactive substances, security of operators and public safety and environment.The above shows that in the planned radiation safety condition described in this report,the practice can be carried out according to the criteria of safety accepted .

Preliminary design study of the TMT Telescope structure system: overview

Science.gov (United States)

Usuda, Tomonori; Ezaki, Yutaka; Kawaguchi, Noboru; Nagae, Kazuhiro; Kato, Atsushi; Takaki, Junji; Hirano, Masaki; Hattori, Tomoya; Tabata, Masaki; Horiuchi, Yasushi; Saruta, Yusuke; Sofuku, Satoru; Itoh, Noboru; Oshima, Takeharu; Takanezawa, Takashi; Endo, Makoto; Inatani, Junji; Iye, Masanori; Sadjadpour, Amir; Sirota, Mark; Roberts, Scott; Stepp, Larry

2014-07-01

We present an overview of the preliminary design of the Telescope Structure System (STR) of Thirty Meter Telescope (TMT). NAOJ was given responsibility for the TMT STR in early 2012 and engaged Mitsubishi Electric Corporation (MELCO) to take over the preliminary design work. MELCO performed a comprehensive preliminary design study in 2012 and 2013 and the design successfully passed its Preliminary Design Review (PDR) in November 2013 and April 2014. Design optimizations were pursued to better meet the design requirements and improvements were made in the designs of many of the telescope subsystems as follows: 1. 6-legged Top End configuration to support secondary mirror (M2) in order to reduce deformation of the Top End and to keep the same 4% blockage of the full aperture as the previous STR design. 2. "Double Lower Tube" of the elevation (EL) structure to reduce the required stroke of the primary mirror (M1) actuators to compensate the primary mirror cell (M1 Cell) deformation caused during the EL angle change in accordance with the requirements. 3. M1 Segment Handling System (SHS) to be able to make removing and installing 10 Mirror Segment Assemblies per day safely and with ease over M1 area where access of personnel is extremely difficult. This requires semi-automatic sequence operation and a robotic Segment Lifting Fixture (SLF) designed based on the Compliance Control System, developed for controlling industrial robots, with a mechanism to enable precise control within the six degrees of freedom of position control. 4. CO2 snow cleaning system to clean M1 every few weeks that is similar to the mechanical system that has been used at Subaru Telescope. 5. Seismic isolation and restraint systems with respect to safety; the maximum acceleration allowed for M1, M2, tertiary mirror (M3), LGSF, and science instruments in 1,000 year return period earthquakes are defined in the requirements. The Seismic requirements apply to any EL angle, regardless of the

Preliminary Study of Single-Phase Natural Circulation for Lab-scaled Molten Salt Application

Energy Technology Data Exchange (ETDEWEB)

Shin, Yukyung; Kang, Sarah; Kim, In Guk; Seo, Seok Bin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Park, Seong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Advanced reactors such as MSR (FHR), VHTR and AHTR utilized molten salt as a coolant for efficiency and safety which has advantages in higher heat capacity, lower pumping power and scale compared to liquid metal. It becomes more necessary to study on the characteristics of molten salt. However, due to several characteristics such as high operating temperature, large-scale facility and preventing solidification, satisfying that condition for study has difficulties. Thus simulant fluid was used with scaling method for lab-scale experiment. Scaled experiment enables simulant fluid to simulate fluid mechanics and heat transfer behavior of molten salt on lower operating temperature and reduced scale. In this paper, as a proof test of the scaled experiment, simplified single-phase natural circulation loop was designed in a lab-scale and applied to the passive safety system in advanced reactor in which molten salt is considered as a major coolant of the system. For the application of the improved safety system, prototype was based on the primary loop of the test-scale DRACS, the main passive safety system in FHR, developed at the OSU. For preliminary experiment, single-phase natural circulation under low power was performed. DOWTHERM A and DOWTHERM RP were selected as simulant candidates. Then, study of feasibility with simulant was conducted based on the scaling law for heat transfer characteristics and geometric parameters. Additionally, simulation with MARS code and ANSYS-CFX with the same condition of natural circulation was carried out as verification. For the accurate code simulation, thermo-physical properties of DOWTHERM A and RP were developed and implemented into MARS code. In this study, single-phase natural circulation experiment was performed with simulant oil, DOWTHERM RP, based on the passive safety system of FHR. Feasibility of similarity experiment for molten salt with oil simulant was confirmed by scaling method. In addition, simulation with two

Los Alamos National Laboratory corregated metal pipe saw facility preliminary safety analysis report. Volume I

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-09-19

This Preliminary Safety Analysis Report addresses site assessment, facility design and construction, and design operation of the processing systems in the Corrugated Metal Pipe Saw Facility with respect to normal and abnormal conditions. Potential hazards are identified, credible accidents relative to the operation of the facility and the process systems are analyzed, and the consequences of postulated accidents are presented. The risk associated with normal operations, abnormal operations, and natural phenomena are analyzed. The accident analysis presented shows that the impact of the facility will be acceptable for all foreseeable normal and abnormal conditions of operation. Specifically, under normal conditions the facility will have impacts within the limits posted by applicable DOE guidelines, and in accident conditions the facility will similarly meet or exceed the requirements of all applicable standards. 16 figs., 6 tabs.

High-intensity focused ultrasound treatment of placenta accreta after vaginal delivery: a preliminary study.

Science.gov (United States)

Bai, Y; Luo, X; Li, Q; Yin, N; Fu, X; Zhang, H; Qi, H

2016-04-01

To evaluate the safety and efficiency of high-intensity focused ultrasound (HIFU) in the treatment of placenta accreta after vaginal delivery. Enrolled into this study between September 2011 and September 2013 were 12 patients who had been diagnosed with placenta accreta following vaginal delivery and who had stable vital signs. All patients were treated using an ultrasound-guided HIFU treatment system. As indication of the effectiveness of the treatment we considered decreased vascular index on color Doppler imaging, decrease in size of residual placenta compared with pretreatment size on assessment by three-dimensional ultrasound with Virtual Organ Computer-aided Analysis, reduced signal intensity and degree of enhancement on magnetic resonance imaging and avoidance of hysterectomy following treatment. To assess the safety of HIFU treatment, we recorded side effects, hemorrhage, infection, sex steroid levels, return of menses and subsequent pregnancy. Patients were followed up in this preliminary study until December 2013. The 12 patients receiving HIFU treatment had an average postpartum hospital stay of 6.8 days and an average period of residual placental involution of 36.9 days. HIFU treatment did not apparently increase the risk of infection or hemorrhage and no patient required hysterectomy. In all patients menstruation recommenced after an average of 80.2 days, and sex steroid levels during the middle luteal phase of the second menstrual cycle were normal. Two patients became pregnant again during the follow-up period. This preliminary study suggests that ultrasound-guided HIFU is a safe and effective non-invasive method to treat placenta accreta patients after vaginal delivery who have stable vital signs and desire to preserve fertility. Copyright © 2015 ISUOG. Published by John Wiley & Sons Ltd. Copyright © 2015 ISUOG. Published by John Wiley & Sons Ltd.

Recommendations: Procedure to develop a preliminary safety report as part of the radioactive waste repository construction licensing process

International Nuclear Information System (INIS)

2003-01-01

The structure of a preliminary safety report for the title purpose should be as follows: A. Textual part: 1. General (Introduction, Basic information about the construction, Timetable); 2. Site information (Siting, Geography and demography, Meteorology and climatic situation, Hydrology, Geology and hydrogeology); 3. Repository design description (Basic function and performance requirements, Design, Auxiliary systems, Fire prevention/protection, Emergency plans); 4. Operation of the repository (Waste acceptance and inspection, Waste handling and interim storage, Waste disposal, Operating monitoring), 5. Health and environmental impact assessment (Radionuclide inventory, Radionuclide transport paths and mechanisms of release into the environment, Radionuclide release in normal and emergency situations, Radiation protection - health impact assessment and regulatory compliance, Draft operating limits and conditions, Proposed ways of assuring physical protection, Uncertainty assessment), 6. Safe repository shutdown/decommissioning concept, 7 Quality assurance assessment, 8. List of selected equipment. B. Annexes: Maps, Drawings, Diagrams, Miscellaneous; C. Documentation: Previous safety report amendments, Protocols, Miscellaneous. (P.A.)

Automating the Generation of Heterogeneous Aviation Safety Cases

Science.gov (United States)

Denney, Ewen W.; Pai, Ganesh J.; Pohl, Josef M.

2012-01-01

A safety case is a structured argument, supported by a body of evidence, which provides a convincing and valid justification that a system is acceptably safe for a given application in a given operating environment. This report describes the development of a fragment of a preliminary safety case for the Swift Unmanned Aircraft System. The construction of the safety case fragment consists of two parts: a manually constructed system-level case, and an automatically constructed lower-level case, generated from formal proof of safety-relevant correctness properties. We provide a detailed discussion of the safety considerations for the target system, emphasizing the heterogeneity of sources of safety-relevant information, and use a hazard analysis to derive safety requirements, including formal requirements. We evaluate the safety case using three classes of metrics for measuring degrees of coverage, automation, and understandability. We then present our preliminary conclusions and make suggestions for future work.

Nuclear safety culture evaluation model based on SSE-CMM

International Nuclear Information System (INIS)

Yang Xiaohua; Liu Zhenghai; Liu Zhiming; Wan Yaping; Peng Guojian

2012-01-01

Safety culture, which is of great significance to establish safety objectives, characterizes level of enterprise safety production and development. Traditional safety culture evaluation models emphasis on thinking and behavior of individual and organization, and pay attention to evaluation results while ignore process. Moreover, determining evaluation indicators lacks objective evidence. A novel multidimensional safety culture evaluation model, which has scientific and completeness, is addressed by building an preliminary mapping between safety culture and SSE-CMM's (Systems Security Engineering Capability Maturity Model) process area and generic practice. The model focuses on enterprise system security engineering process evaluation and provides new ideas and scientific evidences for the study of safety culture. (authors)

Preliminary analysis of K-DEMO thermal hydraulic system using MELCOR; Parametric study of hydrogen explosion

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Lim, Soo Min; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2016-10-15

K-DEMO (Korean fusion demonstration reactor) is future reactor for the commercializing the fusion power generation. The Design of K-DEMO is similar to that of ITER but the fusion energy generation is much bigger because ITER is experimental reactor. For this reason, K-DEMO uses more fusion reaction with bigger amount of tritium. Higher fusion power means more neutron generation that can irradiate the structure around fusion plasma. Fusion reactor can produce many kinds of radioactive material in the accident. Because of this hazard, preliminary safety analysis is mandatory before its construction. Concern for safety problem of accident of fusion/fission reactor has been growing after Fukushima accident which is severe accident from unexpected disaster. To model the primary heat transfer system, in this study, MARS-KS thermal hydraulic analysis is referred. Lee et al. and Kim et al. conducted thermal hydraulic analysis using MARS-KS and multiple module simulation to deal with the phenomena of first wall corrosion for each plasma pulse. This study shows the relationship between vacuum vessel rupture area and source term leakage after hydrogen explosion. For the conservative study, first wall heating is not terminated because the heating inside the vacuum vessel increase the pressure inside VV. Pressurizer, steam generator and turbine is not damaged. 6.69 kg of tritiated water (HTO) and 1 ton of dust is modeled which is ITER guideline. The entire system of K-DEMO is smaller than that of ITER. For this reason, lots of aerosol is release into environment although the safety system like DS is maintained. This result shows that the safety system of K-DEMO should use much more safety system.

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

The quality improvement attitude survey: Development and preliminary psychometric characteristics.

Science.gov (United States)

Dunagan, Pamela B

2017-12-01

To report the development of a tool to measure nurse's attitudes about quality improvement in their practice setting and to examine preliminary psychometric characteristics of the Quality Improvement Nursing Attitude Scale. Human factors such as nursing attitudes of complacency have been identified as root causes of sentinel events. Attitudes of nurses concerning use of Quality and Safety Education for nurse's competencies can be most challenging to teach and to change. No tool has been developed measuring attitudes of nurses concerning their role in quality improvement. A descriptive study design with preliminary psychometric evaluation was used to examine the preliminary psychometric characteristics of the Quality Improvement Nursing Attitude Scale. Registered bedside clinical nurses comprised the sample for the study (n = 57). Quantitative data were analysed using descriptive statistics and Cronbach's alpha reliability. Total score and individual item statistics were evaluated. Two open-ended items were used to collect statements about nurses' feelings regarding their experience in quality improvement efforts. Strong support for the internal consistency reliability and face validity of the Quality Improvement Nursing Attitude Scale was found. Total scale scores were high indicating nurse participants valued Quality and Safety Education for Nurse competencies in practice. However, item-level statistics indicated nurses felt powerless when other nurses deviate from care standards. Additionally, the sample indicated they did not consistently report patient safety issues and did not have a feeling of value in efforts to improve care. Findings suggested organisational culture fosters nurses' reporting safety issues and feeling valued in efforts to improve care. Participants' narrative comments and item analysis revealed the need to generate new items for the Quality Improvement Nursing Attitude Scale focused on nurses' perception of their importance in quality and

Preliminary considerations on safety of computerized control rooms

International Nuclear Information System (INIS)

Vittet, J.

1983-02-01

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

A preliminary study on the application of system dynamics methodology to organizational safety in nuclear power plants: Learning from past models

Energy Technology Data Exchange (ETDEWEB)

Do, Giang [Sol Bridge International School of Business, Daejeon (Korea, Republic of); Kim, Sakil; Lee, Yong Hee; Lee, Yong Hee [KAERI, Daejeon (Korea, Republic of)

2012-10-15

Besides technical design, organizational and human factor are of increasing interest in literature on nuclear safety. Among the methodologies employed to study these factors, System Dynamics (SD) is considered to be suitable for addressing the complexity and dynamicity of the organizational system in nuclear power plants (NPPs). In the following sections, the method will be described and its several prior applications to studying organizational safety will be introduced. An SD model with emphasis on the role of organizational learning in organizational safety will be presented.

Preliminary hazards analysis -- vitrification process

International Nuclear Information System (INIS)

Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P.

1994-06-01

This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility's construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment

Preliminary hazards analysis -- vitrification process

Energy Technology Data Exchange (ETDEWEB)

Coordes, D.; Ruggieri, M.; Russell, J.; TenBrook, W.; Yimbo, P. [Science Applications International Corp., Pleasanton, CA (United States)

1994-06-01

This paper presents a Preliminary Hazards Analysis (PHA) for mixed waste vitrification by joule heating. The purpose of performing a PHA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PHA is then followed by a Preliminary Safety Analysis Report (PSAR) performed during Title 1 and 2 design. The PSAR then leads to performance of the Final Safety Analysis Report performed during the facility`s construction and testing. It should be completed before routine operation of the facility commences. This PHA addresses the first four chapters of the safety analysis process, in accordance with the requirements of DOE Safety Guidelines in SG 830.110. The hazards associated with vitrification processes are evaluated using standard safety analysis methods which include: identification of credible potential hazardous energy sources; identification of preventative features of the facility or system; identification of mitigative features; and analyses of credible hazards. Maximal facility inventories of radioactive and hazardous materials are postulated to evaluate worst case accident consequences. These inventories were based on DOE-STD-1027-92 guidance and the surrogate waste streams defined by Mayberry, et al. Radiological assessments indicate that a facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous materials assessment indicates that a Mixed Waste Vitrification facility will be a Low Hazard facility having minimal impacts to offsite personnel and the environment.

Safety indicators: an efficient tool for a better safety

International Nuclear Information System (INIS)

Aufort, P.; Lars, R.

1993-01-01

Safety indicators based on the examination of the Operating Technical Specifications have been defined with the aim of following the in-operation safety level of French nuclear power plants. These safety indicators are operation feedback tools which permit the a posteriori justification and the adjustment of actual procedures. They would allow detection of an abnormal unavailability occurrence rate or a situation revealing a potential safety problem. So, data acquisition, processing, analysis and display software allowing trend analysis of these indicators has been developed so far as: a reflexion tool for the power plant operators about the safety instructions and the adjustment of preventive maintenance, and a help for decision making at a national level for the examination and the improvement of Operating Technical Specifications. This paper presents the objectives of these safety indicators, the processing tool associated, the preliminary results obtained and more elaborate processing of these indicators. These safety indicators may be very useful in framing probabilistic safety assessments. (author)

Identification of potential safety-related incidents applicable to a breeder fuel reprocessing plant

International Nuclear Information System (INIS)

Perkins, W.C.

1980-01-01

The current emphasis on safety in all phases of the nuclear fuel cycle requires that safety features be identified and included in designs of nuclear facilities at the earliest possible stage. A popular method for the early identification of these safety features is the Preliminary Hazards Analysis. An extension of this analysis is to illustrate the nature of a hazard by its effects in accident situations, that is, to identify what are called safety-related incidents. Some useful tools are described which have been used at the Savannah River Laboratory, SRL, to make Preliminary Hazards Analyses as well as safety analyses of facilities for processing spent nuclear fuels from both power and production reactors. These tools have also been used in safety studies of waste handling operations at the Savannah River Plant. The tools are the SRL Incidents Data Bank and the What If meeting. The application of this methodology to a proposed facility which has breeder fuel reprocessing capability, the Hot Experimental Facility (HEF) is illustrated

Preliminary report on safety aspects on nuclear power generation in Sri Lanka

International Nuclear Information System (INIS)

Jayamanne, D.; Fernando, W.L.W.; Ariyadasa

1988-01-01

This document is intended as background information on nuclear energy to contribute to Sri Lanka's comparative study of alternative sources of energy. This study has considered the safety and environmental effects of nuclear power reactors. Basic concepts of nuclear physics are introduced and providing and appreciation of safety considerations and safety aspects of nuclear power plants and the personnel. Radioactive waste management, storage and disposal are also discussed. Natural radiation levels in Sri Lanka are provided as well as information on biological effects of radiation especially occupational exposure licensing procedures for nuclear power plants are outlined strategy for public awareness of nuclear power is proposed

Immunogenicity and Safety of the New Inactivated Quadrivalent Influenza Vaccine Vaxigrip Tetra: Preliminary Results in Children ≥6 Months and Older Adults

Directory of Open Access Journals (Sweden)

Emanuele Montomoli

2018-03-01

Full Text Available Since the mid-1980s, two lineages of influenza B viruses have been distinguished. These can co-circulate, limiting the protection provided by inactivated trivalent influenza vaccines (TIVs. This has prompted efforts to formulate quadrivalent influenza vaccines (QIVs, to enhance protection against circulating influenza B viruses. This review describes the results obtained from seven phase III clinical trials evaluating the immunogenicity, safety, and lot-to-lot consistency of a new quadrivalent split-virion influenza vaccine (Vaxigrip Tetra® formulated by adding a second B strain to the already licensed TIV. Since Vaxigrip Tetra was developed by means of a manufacturing process strictly related to that used for TIV, the data on the safety profile of TIV are considered supportive of that of Vaxigrip Tetra. The safety and immunogenicity of Vaxigrip Tetra were similar to those of the corresponding licensed TIV. Moreover, the new vaccine elicits a superior immune response towards the additional strain, without affecting immunogenicity towards the other three strains. Vaxigrip Tetra is well tolerated, has aroused no safety concerns, and is recommended for the active immunization of individuals aged ≥6 months. In addition, preliminary data confirm its immunogenicity and safety even in children aged 6–35 months and its immunogenicity in older subjects (aged 66–80 years.

Clinical pharmacokinetics, safety, and preliminary efficacy evaluation of icotinib in patients with advanced non-small cell lung cancer.

Science.gov (United States)

Liu, Dongyang; Zhang, Li; Wu, Yiwen; Jiang, Ji; Tan, Fenlai; Wang, Yingxiang; Liu, Yong; Hu, Pei

2015-09-01

To receive pharmacokinetics, safety, and anti-tumor activity of icotinib, a novel epidermal growth factor receptor (EGFR)-tyrosine kinase inhibitor (TKI), in patients with advanced non-small-cell lung cancer (NSCLC). Patients (n=40) with advanced NSCLC were enrolled to receive escalating doses of icotinib, which was administrated on Day 1 followed by 28-day continuous dosing starting from Day 4. Four dosing regimens, 100mg b.i.d., 150 mg b.i.d., 125 mg t.i.d., and 200mg b.i.d. were studied. Pharmacokinetics (PK), safety, and efficacy of icotinib were evaluated. Icotinib was well tolerated in Chinese patients with refractory NSCLC. No toxicity with >3 grades were reported in more than 2 patients under any dose levels. One complete response (3%) and 9 partial responses (23%) were received. Total disease control rate could reach at 73% and median progress-free survival (range) was 154 (17-462) days. PK exposure of icotinib increased with increase of dose in NSCLC patients. Food was suggested to increase PK exposure by ∼30%. Mean t1/2β was within 5.31-8.07 h. No major metabolite (>10% plasma exposure of icotinib) was found in NSCLC patients. Icotinib with up to 400 mg/day exhibited good tolerance and preliminary antitumor activity in Chinese NSCLC patients. Pharmacokinetics of icotinib and 5 major metabolites were fully investigated in NSCLC patients. Optimized biologic dose (OBD) was finally recommended to be 125 mg t.i.d. for the later clinical study. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

Preliminary safety evaluation for the Simpevarp subarea. Based on data and site descriptions after the initial site investigation stage

International Nuclear Information System (INIS)

2005-04-01

The main objectives of this Preliminary safety evaluation (PSE) of the Simpevarp subarea are: to determine, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB in the report SKB-TR--00-12. These criteria both concern properties of the site judged to be necessary for safety and engineering (requirements) and properties judged to be beneficial (preferences). The findings are then evaluated in order to provide feedback to continued investigations and design work. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The evaluation shows that even considering remaining uncertainties, the Simpevarp subarea meets all safety requirements and most of the safety preferences. Consequently, from a safety point of view, there is no reason not to continue the Site Investigations of the Simpevarp subarea. There are still uncertainties to resolve and the safety would eventually need to be verified through a full safety assessment. Still, this Preliminary Safety Evaluation demonstrates that it is likely that a safe repository for spent nuclear fuel of the KBS-3 type could be constructed at the site. The following feedback is provided to the site investigations and the associated site modelling: Reducing the uncertainty on the deformation zone geometry within the Simpevarp subarea would allow for a more specified layout, although the sensitivity analysis shows that the space needed is rather robust with respect to uncertainties in the zones. There is substantial uncertainty in the discrete fracture network (DFN) model

Learning from positively deviant wards to improve patient safety: an observational study protocol.

Science.gov (United States)

Baxter, Ruth; Taylor, Natalie; Kellar, Ian; Lawton, Rebecca

2015-12-11

Positive deviance is an asset-based approach to improvement which has recently been adopted to improve quality and safety within healthcare. The approach assumes that solutions to problems already exist within communities. Certain groups or individuals identify these solutions and succeed despite having the same resources as others. Within healthcare, positive deviance has previously been applied at individual or organisational levels to improve specific clinical outcomes or processes of care. This study explores whether the positive deviance approach can be applied to multidisciplinary ward teams to address the broad issue of patient safety among elderly patients. Preliminary work analysed National Health Service (NHS) Safety Thermometer data from 34 elderly medical wards to identify 5 'positively deviant' and 5 matched 'comparison' wards. Researchers are blinded to ward status. This protocol describes a multimethod, observational study which will (1) assess the concurrent validity of identifying positively deviant elderly medical wards using NHS Safety Thermometer data and (2) generate hypotheses about how positively deviant wards succeed. Patient and staff perceptions of safety will be assessed on each ward using validated surveys. Correlation and ranking analyses will explore whether this survey data aligns with the routinely collected NHS Safety Thermometer data. Staff focus groups and researcher fieldwork diaries will be completed and qualitative thematic content analysis will be used to generate hypotheses about the strategies, behaviours, team cultures and dynamics that facilitate the delivery of safe patient care. The acceptability and sustainability of strategies identified will also be explored. The South East Scotland Research Ethics Committee 01 approved this study (reference: 14/SS/1085) and NHS Permissions were granted from all trusts. Findings will be published in peer-reviewed, scientific journals, and presented at academic conferences. This study

Safety impacts of bicycle infrastructure: A critical review.

Science.gov (United States)

DiGioia, Jonathan; Watkins, Kari Edison; Xu, Yanzhi; Rodgers, Michael; Guensler, Randall

2017-06-01

This paper takes a critical look at the present state of bicycle infrastructure treatment safety research, highlighting data needs. Safety literature relating to 22 bicycle treatments is examined, including findings, study methodologies, and data sources used in the studies. Some preliminary conclusions related to research efficacy are drawn from the available data and findings in the research. While the current body of bicycle safety literature points toward some defensible conclusions regarding the safety and effectiveness of certain bicycle treatments, such as bike lanes and removal of on-street parking, the vast majority treatments are still in need of rigorous research. Fundamental questions arise regarding appropriate exposure measures, crash measures, and crash data sources. This research will aid transportation departments with regard to decisions about bicycle infrastructure and guide future research efforts toward understanding safety impacts of bicycle infrastructure. Copyright © 2017 Elsevier Ltd and National Safety Council. All rights reserved.

Preliminary safety analysis of unscrammed events for KLFR

International Nuclear Information System (INIS)

Kim, S.J.; Ha, G.S.

2005-01-01

The report presents the design features of KLFR; Safety Analysis Code; steady-state calculation results and analysis results of unscrammed events. The calculations of the steady-state and unscrammed events have been performed for the conceptual design of KLFR using SSC-K code. UTOP event results in no fuel damage and no centre-line melting. The inherent safety features are demonstrated through the analysis of ULOHS event. Although the analysis of ULOF has much uncertainties in the pump design, the analysis results show the inherent safety characteristics. 6% flow of rated flow of natural circulation is formed in the case of ULOF. In the metallic fuel rod, the cladding temperature is somewhat high due to the low heat transfer coefficient of lead. ULOHS event should be considered in design of RVACS for long-term cooling

Occupational health and safety in the Moroccan construction sites: preliminary diagnosis

Science.gov (United States)

Tarik, Bakeli; Adil, Hafidi Alaoui

2018-05-01

Managing occupational health and safety on Moroccan construction sector represents the first step for projects' success. In fact, by avoiding accidents, all the related direct and indirect costs and delays can be prevented. That leads to an important question always asked by any project manager: what are the factors responsible for accidents? How can they be avoided? Through this research, the aim is to go through these questions, to contribute in occupational health and safety principles understanding, to identify construction accidentology and risk management opportunities and to approach the case of Moroccan construction sites by an accurate diagnosis. The approach is to make researchers, managers, stakeholders and deciders aware about the criticality of construction sites health and safety situation. And, to do the first step for a scientific research project in relation with health and safety in the Moroccan construction sector. For this, the paper will study the related state of art namely about construction sites accidents causation, and will focus on Reason's `Swiss cheese' model and its utilization for Moroccan construction sites health and safety diagnosis. The research will end with an estimation of an accidents fatality rate in the Moroccan construction sector and a benchmarking with the international rates. Finally, conclusions will be presented about the necessity of Occupational Health and Safety Management System (OHSMS) implementation, which shall cover all risk levels, and insure, at the same time, that the necessary defenses against accidents are on place.

Development of a Preliminary Model for Evaluating Occupational Health and Safety Risk Management Maturity in Small and Medium-Sized Enterprises

Directory of Open Access Journals (Sweden)

Bilal Kaassis

2018-02-01

Full Text Available Management of occupational health and safety (OHS risks is a crucial component of any business. Numerous investigations have shown that work-related injuries and deaths occur disproportionately in small-to-medium-sized enterprises (SMEs and that this is clearly due to deficient management of OHS risks. The main goal of this work is to develop a base of indicators suitable for evaluating OHS risk management maturity in industrial SMEs. A preliminary model is then proposed for this evaluation, based on a small number of relevant indicators selected from a careful bibliographic review. The work begins with a critical review of the literature and analysis of known concepts, methods, tools and models of measurement of risk analysis maturity in order to extract relevant indicators. The most suitable indicators are then grouped to form the basis of a preliminary model for evaluating OHS risk management maturity in the SME setting. Our findings will help managers of SMEs make sound decisions in their quest to improve the OHS performance of their businesses.

A two-step approach for the preliminary evaluation of the thermal-hydraulics and safety of the ELSY open square core design

International Nuclear Information System (INIS)

Meloni, Paride; Bandini, Giacomino; Polidori, Massimiliano; Cervone, Antonio; Manservisi, Sandro

2009-01-01

Several innovative solutions for a liquid metal fast reactor design have been investigated in the EURATOM Sixth Framework Programme and an open-assembly core design for the ELSY (European Lead-cooled System) reactor has been proposed by ENEA. The development of this new reactor, based on innovative neutronic and safety considerations, requires a new approach to the thermal-hydraulic (T/H) core design. In this paper a new two-step approach of the T/H analysis for this open-assembly core is presented and, in particular is used for the evaluation of the preliminary core design of a 1500 MW lead fast reactor with open square lattice and three fuel radial zones with different levels of enrichment. In the first step a preliminary thermal-hydraulic and safety evaluation of the core neutronic design is investigated by using a one-dimensional RELAP5 model for independent channel analysis. Then two and three-dimensional effects are taken into account by using a dedicated tool for the evaluation of assembly mixing effects. The RELAP5 model, based on pressure loss and heat transfer correlations available for heavy liquid metal flows in rod bundle, consists of completely independent assemblies and therefore it can be used for a conservative evaluation of the thermal-hydraulics of the core reactor. Due to the open-lattice configuration, the two and three-dimensional effects are important and they are taken into account by using a simplified three-dimensional numerical model of an open square lattice reactor core, developed with the purpose of analyzing the whole core behavior. The numerical simulation is performed at assembly length level taking into account the local fluctuations of turbulent viscosity and energy exchange coefficients at sub-channel level through transfer operators based on parametric coefficients. A preliminary evaluation of the mixing effects between assembly flows on the temperature field has been performed by using an average assembly turbulent viscosity

A Preliminary Review of Fatigue Among Rail Staff

OpenAIRE

Jialin Fan; Andrew P. Smith

2018-01-01

Background: Fatigue is a severe problem in the rail industry, which may jeopardize train crew's health and safety. Nonetheless, a preliminary review of all empirical evidence for train crew fatigue is still lacking. The aim of the present paper is, therefore, to provide a preliminary description of occupational fatigue in the rail industry. This paper reviews the literature with the research question examining the risk factors associated with train crew fatigue, covering both papers published...

Development of safety analysis technology for LMR

International Nuclear Information System (INIS)

Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

2000-05-01

The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

Development of safety analysis technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

2000-05-01

The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

Improving safety culture through the health and safety organization: a case study.

Science.gov (United States)

Nielsen, Kent J

2014-02-01

International research indicates that internal health and safety organizations (HSO) and health and safety committees (HSC) do not have the intended impact on companies' safety performance. The aim of this case study at an industrial plant was to test whether the HSO can improve company safety culture by creating more and better safety-related interactions both within the HSO and between HSO members and the shop-floor. A quasi-experimental single case study design based on action research with both quantitative and qualitative measures was used. Based on baseline mapping of safety culture and the efficiency of the HSO three developmental processes were started aimed at the HSC, the whole HSO, and the safety representatives, respectively. Results at follow-up indicated a marked improvement in HSO performance, interaction patterns concerning safety, safety culture indicators, and a changed trend in injury rates. These improvements are interpreted as cultural change because an organizational double-loop learning process leading to modification of the basic assumptions could be identified. The study provides evidence that the HSO can improve company safety culture by focusing on safety-related interactions. © 2013. Published by Elsevier Ltd and National Safety Council.

Seismic safety of Paks nuclear power plant

International Nuclear Information System (INIS)

Katona, T.

1993-01-01

An extensive program is underway at Paks NPP for evaluation of the seismic safety and for development of the necessary safety increasing measures. This program includes the following five measures: investigation of methods, regulations and techniques utilized for reassessment of seismic safety of operating NPPs and promoting safety; investigation of earthquake hazards; development of concepts for creating the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept, and preliminary evaluation of the seismic safety

Preliminary safety analysis report for the Waste Characterization Facility

International Nuclear Information System (INIS)

1994-10-01

This safety analysis report outlines the safety concerns associated with the Waste Characterization Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are to: define and document a safety basis for the Waste Characterization Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume. 142 refs., 38 figs., 39 tabs

Preliminary safety assessments in construction of the pilot industrial facility for final disposal of low and intermediate radioactive waste in the archipelago Novaya Zemlya

International Nuclear Information System (INIS)

Lopatin, V.V.; Lobanov, N.F.; Mankin, V.I.; Karamushka, V.P.; Ostroborodov, V.V.

1999-01-01

This presentation discusses a preliminary safety evaluation of radioactive waste burial at the experimental plant located on Novaya Zemlya. The issues considered are (1) the main provisions on radioactive waste burial in permafrost rock, (2) mining, geological and geocryological conditions at the experimental works' operating site, (3) the main properties of solid and solidified radioactive wastes, (4) the main parameters of the experimental works, (5) preliminary evaluation of safety. The evaluation includes the main requirements to geocryologic characteristics of the permafrost rock intended for waste burial and analyses the seasonal mining-geological and geocryological conditions in the area of the experimental works. The area is situated within the limits of the southern Novozemelsky anticlinorium composed of the Silurian, Devonian and carboniferous rocks of the Paleozoic group. It is mainly limestone and dolomite, showing in rock sequence the layers, benches and horizons of clay shales, aleurolites, conglomerates and magmatic rocks covered with a thin Quaternary sedimentary mantle on the surface. The area is characterised by a confluent continuous layer no less than 300 m thick, seasonal thawing depth 0.5-2.0 m, annual zero temperature variations 10-15 m by the depth, and mean annual rock temperature of -4.5 - 5.0 C. The plant is an independent enterprise supplied with all the required services for industrial and communal/living purposes. The evaluation studies two possible scenarios for accidents during transport of waste to Novaya Zemlya, and the consequences of damage to the plant caused by the impact of a celestial body/flying object, by a catastrophic earthquake, and the effect of global climate warming in the Arctic area

Preliminary study of mercury target structure

Energy Technology Data Exchange (ETDEWEB)

Kaminaga, Masanori; Haga, Katsuhiro; Hino, Ryutaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kumasaka, Katsuyuki; Uchida, Shoji; Nakagawa, Toshi; Mori, Seiji; Nishikawa, Akira

1997-11-01

Development of a proton accelerator based neutron source (1.5 GeV, 5.3 mA (for neutron source 3.3 mA), thermal power 8 MW) is currently conducted by the Special Task Force for Neutron Science Initiative, JAERI. Preliminary design studies and related R and D of a solid metal target for the first stage (1.5 GeV, 1 mA) and a liquid metal target for both the first and second stages (1.5 GeV, 3.3 mA) are conducted by the Target Group to develop both solid and liquid metal target systems. A few kinds of target structures have been investigated in FY 1996 and the preliminary results for the target structures are described in this paper. Investigation results of alternative materials for the target container are also described in this paper. (author)

Preliminary study to questions relating to the safety of nuclear power plants A and B at Biblis

International Nuclear Information System (INIS)

Fischer, B.; Hahn, L.; Rausch, L.

1985-01-01

With a view to developing suitable tools for the safety evaluation of reactors A and B at Biblis, the publication compiles all aspects relevant to safety, creates an evaluation frame, and evaluates the aspects relevant to safety by means of this frame of evaluation. According to the composition of the work, the overall subject is split up into the complexes information, acquisition, evaluation of operational experience, probabilistic analyses, comparison with newer PWR type reactors, fulfilling of injunctions, modifications due to disposal problems, and the disposal situation. (DG) [de

Intraarterial reteplase and intravenous abciximab for treatment of acute ischemic stroke. A preliminary feasibility and safety study in a non-human primate model

International Nuclear Information System (INIS)

Qureshi, Adnan I.; Suri, M. Fareed K.; Ali, Zulfiqar; Ringer, Andrew J.; Boulos, Alan S.; Guterman, Lee R.; Hopkins, L. Nelson; Nakada, Marian T.; Alberico, Ronald A.; Martin, Lisa B.E.

2005-01-01

We performed a preliminary feasibility and safety study using intravenous (IV) administration of a platelet glycoprotein IIb/IIIa inhibitor (abciximab) in conjunction with intraarterial (IA) administration of a thrombolytic agent (reteplase) in a primate model of intracranial thrombosis. We introduced thrombus through superselective catheterization of the intracranial segment of the internal carotid artery in 16 primates. The animals were randomly assigned to receive IA reteplase and IV abciximab (n =4), IA reteplase and IV placebo (n =4), IA placebo and IV abciximab (n =4) or IA and IV placebo (n =4). Recanalization was assessed by serial angiography during the 6-h period after initiation of treatment. Postmortem magnetic resonance (MR) imaging was performed to determine the presence of cerebral infarction or intracranial hemorrhage. Partial or complete recanalization at 6 h after initiation of treatment (decrease of two or more points in pre-treatment angiographic occlusion grade) was observed in two animals treated with IA reteplase and IV abciximab, three animals treated with IA reteplase alone and one animal treated with IV abciximab alone. No improvement in perfusion was observed in animals that received IV and IA placebo. Cerebral infarction was demonstrated on postmortem MR imaging in three animals that received IA and IV placebo and in one animal each from the groups that received IA reteplase and IV abciximab or IV abciximab alone. One animal that received IV abciximab alone had a small intracerebral hemorrhage on MR imaging. (orig.)

Preliminary Results Of LOCA Problem For APR1400 Reactor

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Le Thi Thu; Vo Thi Huong; Le Van Hong

2011-01-01

Several features of NPP with APR1400 nuclear reactor during a loss of coolant accident (LOCA) are investigated in this study. The report describes some main design characteristics of an engineering safety systems of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Large Break LOCA accident has been analyzed and evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with broken area of 0.0465 m 2 in case of high pressure safety injection system (HPSI) failed to operate or 2 and 4 HPSI pumps are activated. The preliminary results of this work is a part of collaboration between INST researchers and KAERI experts in using RELAP tool for safety analysis of NPPs. (author)

A Preliminary Study on the Measures to Assess the Organizational Safety: The Cultural Impact on Human Error Potential

International Nuclear Information System (INIS)

Lee, Yong Hee; Lee, Yong Hee

2011-01-01

The Fukushima I nuclear accident following the Tohoku earthquake and tsunami on 11 March 2011 occurred after twelve years had passed since the JCO accident which was caused as a result of an error made by JCO employees. These accidents, along with the Chernobyl accident, associated with characteristic problems of various organizations caused severe social and economic disruptions and have had significant environmental and health impact. The cultural problems with human errors occur for various reasons, and different actions are needed to prevent different errors. Unfortunately, much of the research on organization and human error has shown widely various or different results which call for different approaches. In other words, we have to find more practical solutions from various researches for nuclear safety and lead a systematic approach to organizational deficiency causing human error. This paper reviews Hofstede's criteria, IAEA safety culture, safety areas of periodic safety review (PSR), teamwork and performance, and an evaluation of HANARO safety culture to verify the measures used to assess the organizational safety

A Preliminary Study on the Measures to Assess the Organizational Safety: The Cultural Impact on Human Error Potential

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee; Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

The Fukushima I nuclear accident following the Tohoku earthquake and tsunami on 11 March 2011 occurred after twelve years had passed since the JCO accident which was caused as a result of an error made by JCO employees. These accidents, along with the Chernobyl accident, associated with characteristic problems of various organizations caused severe social and economic disruptions and have had significant environmental and health impact. The cultural problems with human errors occur for various reasons, and different actions are needed to prevent different errors. Unfortunately, much of the research on organization and human error has shown widely various or different results which call for different approaches. In other words, we have to find more practical solutions from various researches for nuclear safety and lead a systematic approach to organizational deficiency causing human error. This paper reviews Hofstede's criteria, IAEA safety culture, safety areas of periodic safety review (PSR), teamwork and performance, and an evaluation of HANARO safety culture to verify the measures used to assess the organizational safety

Preliminary Study on Effect of Aviation Fuel in the Safety Evaluation of Nuclear Power Plant Crashed by Aircraft

International Nuclear Information System (INIS)

Jin, Byeong Moo; Jeon, Se Jin; Lee, Yun Seok; Kim, Young Jin

2011-01-01

As the safety assessments of nuclear power plants for the hypothetical large civil aircraft crash should be made mandatory, studies on large aircraft-nuclear power plant impact analyses and assessments are actively in progress. The large civil aircraft are being operated with a large amount of fuel and the fuel can be assumed to contribute to the impact loads at the impact. The fuel, i.e., the internal liquid can be considered as added masses classically in the evaluation of the impact load. According to the recent experimental research, it has been shown that the impact load of high speed impacting body with internal liquid is much higher than that of the mass-equivalent impacting body. In this study, the impact loads according to the existence of the internal liquid are computed by numerical methods and the safety assessment of nuclear power plant crashed by large civil aircraft are performed as an application

Joint SKI and SSI review of SKB preliminary safety assessment of repository for long-lived low- and intermediate-level waste. Review report

International Nuclear Information System (INIS)

2001-03-01

SKI and SSI find that SKB's first proper safety assessment of the SFL 3-5 repositories provides a valuable springboard for continued efforts in this field. Even though the safety assessment is relatively limited in scope, it has numerous merits. The specific problems associated with the chosen repository concept for SFL 3-5 are discussed in a generally transparent manner. On the other hand, the authorities consider that SKB have only partly achieved the expressed goal of studying the significance of the current repository design and the choice of site. The greatest deficiency consists in that neither internal disturbances (such as considerable cracking or degradation of concrete structures) nor external disturbances (such as the effects of climate changes and glaciation) have been addressed in a thorough manner. A coherent report justifying the design choice from a long-term safety perspective is, in large part, not found here. SKI and SSI recommend that SKB provide a comparison with other possible SFL 3-5 repository designs. Depending upon, among other factors, what geospheric and biospheric conditions are assumed, SKB have shown that the calculated dose values could be relatively high for certain cases. More realistic assessments would be needed to draw reasonable comparisons between different sites, and to evaluate the importance of different nuclides in different contexts. Our review of SKBs preliminary safety assessment indicates that a great deal of research and development work remains to be done before the level of knowledge in this field is comparable with that associated with the final repository for spent fuel. This is reflected with unanimity in the international expert committee's review, and in the consultants' reviews. SKI and SSI wish to point out in particular the fact that comparison with SFR is of limited value, since the safety associated with SFL 3- 5 must be assessed on a much longer time scale. SKI and SSI find it remarkable that SKB have

Preliminary Hazards Analysis Plasma Hearth Process

International Nuclear Information System (INIS)

Aycock, M.; Coordes, D.; Russell, J.; TenBrook, W.; Yimbo, P.

1993-11-01

This Preliminary Hazards Analysis (PHA) for the Plasma Hearth Process (PHP) follows the requirements of United States Department of Energy (DOE) Order 5480.23 (DOE, 1992a), DOE Order 5480.21 (DOE, 1991d), DOE Order 5480.22 (DOE, 1992c), DOE Order 5481.1B (DOE, 1986), and the guidance provided in DOE Standards DOE-STD-1027-92 (DOE, 1992b). Consideration is given to ft proposed regulations published as 10 CFR 830 (DOE, 1993) and DOE Safety Guide SG 830.110 (DOE, 1992b). The purpose of performing a PRA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PRA then is followed by a Preliminary Safety Analysis Report (PSAR) performed during Title I and II design. This PSAR then leads to performance of the Final Safety Analysis Report performed during construction, testing, and acceptance and completed before routine operation. Radiological assessments indicate that a PHP facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous material assessments indicate that a PHP facility will be a Low Hazard facility having no significant impacts either onsite or offsite to personnel and the environment

Development and Validation of a Safety Attitude Scale for Coal Miners in China

Directory of Open Access Journals (Sweden)

Xiang Wu

2017-11-01

Full Text Available Safety attitude is of vital importance to accident prevention, and the high accident rate in the coal mining industry makes it urgent to undertake research on coal miners’ safety attitude. However, the current literature still lacks a valid and reliable safety attitude measurement scale for coal miners, which stands as a barrier against their safety attitude improvement. In this study, a scale is developed that can be used to measure coal miners’ safety attitude. The preliminary scale was based on an extended literature review. Empirical data were then collected from 725 coal miners using the preliminary scale. Both exploratory and confirmatory factor analyses were undertaken to validate and improve the scale. The final scale, which consists of 17 items, contains four dimensions: management safety commitment, team safety climate, fatalism and work pressure. Results show that this safety attitude scale can effectively measure the safety attitude of coal miners, showing high psychological measurement validity. This paper contributes to the occupational safety research by developing the factor structure and indicator system of coal miners’ safety attitude, thus providing more profound interpretation of this crucial construct in the safety research domain. The measurement scale serves as an important tool for safety attitude benchmarking among different coal mining enterprises and, thus, can boost the overall safety improvement of the whole industry. These findings can facilitate improvement of both theories and practices related to occupational safety attitude.

Safety and tolerability of transcranial direct current stimulation to stroke patients - A phase I current escalation study.

Science.gov (United States)

Chhatbar, Pratik Y; Chen, Rong; Deardorff, Rachael; Dellenbach, Blair; Kautz, Steven A; George, Mark S; Feng, Wuwei

A prior meta-analysis revealed that higher doses of transcranial direct current stimulation (tDCS) have a better post-stroke upper-extremity motor recovery. While this finding suggests that currents greater than the typically used 2 mA may be more efficacious, the safety and tolerability of higher currents have not been assessed in stroke patients. We aim to assess the safety and tolerability of single session of up to 4 mA in stroke patients. We adapted a traditional 3 + 3 study design with a current escalation schedule of 1»2»2.5»3»3.5»4 mA for this tDCS safety study. We administered one 30-min session of bihemispheric montage tDCS and simultaneous customary occupational therapy to patients with first-ever ischemic stroke. We assessed safety with pre-defined stopping rules and investigated tolerability through a questionnaire. Additionally, we monitored body resistance and skin temperature in real-time at the electrode contact site. Eighteen patients completed the study. The current was escalated to 4 mA without meeting the pre-defined stopping rules or causing any major safety concern. 50% of patients experienced transient skin redness without injury. No rise in temperature (range 26°C-35 °C) was noted and skin barrier function remained intact (i.e. body resistance >1 kΩ). Our phase I safety study supports that single session of bihemispheric tDCS with current up to 4 mA is safe and tolerable in stroke patients. A phase II study to further test the safety and preliminary efficacy with multi-session tDCS at 4 mA (as compared with lower current and sham stimulation) is a logical next step. ClinicalTrials.gov Identifier: NCT02763826. Copyright © 2017 Elsevier Inc. All rights reserved.

Survey of systems safety analysis methods and their application to nuclear waste management systems

International Nuclear Information System (INIS)

Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

1981-11-01

This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study

Survey of systems safety analysis methods and their application to nuclear waste management systems

Energy Technology Data Exchange (ETDEWEB)

Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

1981-11-01

This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

Preliminary safety analysis for key design features of KALIMER

Energy Technology Data Exchange (ETDEWEB)

Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

2000-07-01

KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

Preliminary studies in rice-fish culture in a rainfed lowland ecology ...

African Journals Online (AJOL)

Preliminary studies in rice-fish culture in a rainfed lowland ecology in Ghana. PKA Dartey, RK Bam, J Ofori. Abstract. Mixed farms of rice and fish are yet to receive attention in Ghana, despite lowland rice being grown under inundation in most areas nationwide. In a preliminary study, Nile tilapia (Oreochromis niloticus) was ...

Anterior capsular staining with trypan blue for capsulorhexis in mature and hypermature cataracts. A preliminary study

Directory of Open Access Journals (Sweden)

Kothari Kulin

2001-01-01

Full Text Available Purpose: To study the efficacy and safety of 0.1% Trypan Blue dye to stain the anterior capsule for capsulorhexis in mature and hypermature cataracts. Methods: This preliminary study included 25 eyes of 25 patients with a unilateral mature or hypermature cataract, including one case of traumatic mature cataract. In all these cases 0.2ml of 0.1% trypan blue dye was used to stain the anterior capsule. The efficacy and safety of the dye was evaluated on the basis of intraoperative and postoperative observations. Results: In all 25 eyes the capsulorhexis was completed. There was peripheral extension of the capsulorhexis in the eye with traumatic cataract and the stained edge of the anterior capsule helped identification and redirection of the capsulorhexis. Successful phacoemulsification with intraocular lens implantation was performed in all eyes. Adverse reactions related to the dye such as raised intraocular pressure, anterior chamber inflammation and endothelial damage were not observed in the immediate postoperative period or at the end of mean follow-up of 3 months. Conclusion: Trypan blue dye staining of the anterior capsule appears to be a very useful and safe technique that simplifies capsulorhexis in mature and hypermature cataracts.

New Reactor Siting in Finland, Hanhikivi Site in Pyhaejoki - STUK preliminary safety assessment

International Nuclear Information System (INIS)

Nevalainen, Janne

2013-01-01

STUK has performed a preliminary assessment of the Decision-in-Principle on the Fennovoima application. A variety of factors must be considered in the selection of a site, including effects of the site on the plant design and the effects of the plant on the site environment. These include external hazards, both natural and human-induced. Since this is a new site, an extensive siting process is followed, that can include an EIA. A site survey is performed to identify candidate sites, after investigating a large region and rejecting unsuitable sites. The remaining sites are then screened and compared on the basis of safety and other considerations to select one or more preferred sites. Natural hazards include geology, seismology, hydrology and meteorology. Offshore ice will be a particular hazard for this plant, since the site is on average only 1.5 m above sea level. The design basis earthquake corresponds to a return frequency of 100,000 years, with 50 % confidence. The existing sites in southern Finland used a design peak ground acceleration of 0.1 g with the ground response spectrum maximum at 10 Hz. The candidate sites in northern Finland will require a peak ground acceleration of 0.2 g with the ground response spectrum maximum at 25 Hz

Pharmacognostic standardization and preliminary phytochemical studies of Gaultheria trichophylla.

Science.gov (United States)

Alam, Fiaz; Najum us Saqib, Qazi

2015-01-01

Gaultheria trichophylla Royle (Ericaceae) has long been used for various ailments in traditional systems of medicines; most importantly it is used against pain and inflammation. This study determines various pharmacognostic and phytochemical standards helpful to ensure the purity, safety, and efficacy of medicinal plant G. trichophylla. Intact aerial parts, powdered materials, and extracts were examined macro- and microscopically and pharmacognostic standardization parameters were determined in accordance with the guidelines given by the World Health Organization (WHO). Parameters including extractive values, ash values, and loss on drying were determined. Preliminary phytochemical tests, fluorescence analysis, and chromatographic profiling were performed for the identification and standardization of G. trichophylla. The shape, size, color, odor, and surface characteristics were noted for intact drug and powdered drug material of G. trichophylla. Light and scanning electron microscope images of cross section of leaf and powdered microscopy revealed useful diagnostic features. Histochemical, phytochemical, physicochemical, and fluorescence analysis proved useful tools to differentiate the powdered drug material. High-performance liquid chromatography (HPLC) analysis showed the presence of important phytoconstituents such as gallic acid, rutin, and quercetin. The data generated from the present study help to authenticate the medicinally important plant G. trichophylla. Qualitative and quantitative microscopic features may be helpful for establishing the pharmacopeia standards. Morphology as well as various pharmacognostic aspects of different parts of the plant were studied and described along with phytochemical and physicochemical parameters, which could be helpful in further isolation and purification of medicinally important compounds.

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

International Nuclear Information System (INIS)

Kirk, W.L.

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

Energy Technology Data Exchange (ETDEWEB)

Kirk, W.L. (comp.)

1976-08-01

The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

Safety of high speed ground transportation systems: X2000 US demonstration vehicle dynamics trials, preliminary test report. Report for October 1992-January 1993

Energy Technology Data Exchange (ETDEWEB)

Whitten, B.T.; Kesler, J.K.

1993-01-01

The report documents the procedures, events, and results of vehicle dynamic tests carried out on the ASEA-Brown Boveri (ABB) X2000 tilt body trainset in the US between October 1992 and January 1993. These tests, sponsored by Amtrak and supported by the FRA, were conducted to assess the suitability of the X2000 trainset for safe operation at elevated cant deficiencies and speeds in Amtrak's Northeast Corridor under existing track conditions in a revenue service demonstration. The report describes the safety criteria against which the performance of the X2000 test train was examined, the instrumentation used, the test locations, and the track conditions. Preliminary results are presented from tests conducted on Amtrak lines between Philadelphia and Harrisburg, PA, and between Washington DC and New York NY, in which cant deficiencies of 12.5 inches and speeds of 154 mph were reached in a safe and controlled manner. The significance of the results is discussed, and preliminary conclusions and recommendations are presented.

Preliminary waste acceptance requirements for the planned Konrad repository

International Nuclear Information System (INIS)

Warnecke, E.; Brennecke, P.

1987-01-01

The Physikalisch-Technische Bundesanstalt (PTB) has established Preliminary Waste Acceptance Requirements for the planned Konrad repository. These requirements were developed, in accordance with the Safety Criteria of the Reactor Safety Commission, with the help of a site specific safety assessment; they are under the reservation of the plan approval procedure, which is still in progress. In developing waste acceptance requirements, the PTB fulfills one of its duties as the institute responsible for waste disposal and gives guidelines for waste conditioning to waste producers and conditioners. (orig.) [de

Preliminary I&C Design for LORELEI

International Nuclear Information System (INIS)

Korotkin, S.; Kaufman, Y.; Guttmann, E. B.; Levy, S.; Amidan, D.; Gdalyho, B.; Cahana, T.; Ellenbogen, A.; Arad, M.; Weiss, Y.; Sasson, A.; Ferry, L.; Bourrelly, F.; Cohen, Y.

2014-01-01

This document summarizes the preliminary I&C design for LORELEI experiment The preliminary design deals with considerations regarding appropriate safety and service instrumentation. The determined closed loop control rules for temperature and position will be implemented in the detailed design. The Computer Aided Operator Decisions System (CAODS) will be used for prediction of hot spot temperature and thickness of oxidation layer using Baker-Just correlation. The proposed hybrid simulation system comprising of both virtual and real hardware will be in-cooperated for LORELEI verification. It will perform both integration cold tests for a partial hardware loop and virtual tests for the final I&C design

[Development and validation of the Korean patient safety culture scale for nursing homes].

Science.gov (United States)

Yoon, Sook Hee; Kim, Byungsoo; Kim, Se Young

2013-06-01

The purpose of this study was to develop a tool to evaluate patient safety culture in nursing homes and to test its validity and reliability. A preliminary tool was developed through interviews with focus group, content validity tests, and a pilot study. A nationwide survey was conducted from February to April, 2011, using self-report questionnaires. Participants were 982 employees in nursing homes. Data were analyzed using Cronbach's alpha, item analysis, factor analysis, and multitrait/multi-Item analysis. From the results of the analysis, 27 final items were selected from 49 items on the preliminary tool. Items with low correlation with total scale were excluded. The 4 factors sorted by factor analysis contributed 63.4% of the variance in the total scale. The factors were labeled as leadership, organizational system, working attitude, management practice. Cronbach's alpha for internal consistency was .95 and the range for the 4 factors was from .86 to .93. The results of this study indicate that the Korean Patient Safety Culture Scale has reliability and validity and is suitable for evaluation of patient safety culture in Korean nursing homes.

Criteria for safety-related nuclear plant operator actions: a preliminary assessment of available data

International Nuclear Information System (INIS)

Haas, P.M.; Bott, T.F.

1980-01-01

In the US, an effort has been underway for a number of years to develop a design standard to define when required manual operator action can be accepted as part of a nuclear plant design basis. Insufficient data are available to provide quantitative guidelines for the standard. To provide the necessary data base to support such standards and the necessary quantitative assessment of operator reliability, the US Nuclear Regulatory Commission is sponsoring a study at Oak National Laboratory to develop the data base. A preliminary assessment completed in April, 1979 concluded that sufficient data from US operating experience did not exist to provide an adequate data base. A program of research using full-scope nuclear plant simulators and results that are correlated to field data was suggested. That program was recently initiated. The approach, results and conclusions of the preliminary assessment are reviewed and the planned research program of simulator studies is summarised. (author)

Nordic studies in reactor safety

International Nuclear Information System (INIS)

Pershagen, N.

1993-01-01

The Nordic Nuclear Safety Research Programme SIK programme in reactor safety is part of a major joint Nordic research effort in nuclear safety. The report summarizes the achievements of the SIK programme, which was carried out during 1990-1993 in collaboration between Nordic nuclear utilities, safety authorities, and research institutes. Three main projects were successfully completed dealing with: 1) development and application of a living PSA concept for monitoring the risk of core damage, and of safety indicators for early warning of possible safety problems; 2) review and intercomparison of severe accident codes, case studies of potential core melt accidents in nordic reactors, development of chemical models for the MAAP code, and outline of a system for computerized accident management support; 3) compilation of information about design and safety features of neighbouring reactors in Germany, Lithuania and Russia, and for naval reactors and nuclear submarines. The report reviews the state-of-the-art in each subject matter as an introduction to the individual project summaries. The main findings of each project are highlighted. The report also contains an overview of reactor safety research in the Nordic countries and a summary of fundamental reactor safety principles. (au) (69 refs.)

Experimental method and preliminary studies of the passive containment water film evaporation mass transfer

Energy Technology Data Exchange (ETDEWEB)

Li, Cheng [State Nuclear Power Technology Research, Beijing (China). Development Center; State Nuclear Power Research Institute, Beijing (China); Yang, Lin; Zhao, Wei; Zhou, Shan; Du, Wangfang; Gao, Zhan; Li, Honegsen [State Nuclear Power Technology Research, Beijing (China). Development Center

2017-05-15

For larger containments and higher operation parameters, characteristics of the outside cooling of the PCCS are very important for the analysis on the containment integrity. A preliminary analysis was made and a four-step experimental method was used to numerically analyze the falling water film evaporation for the advanced passive containment. Then, the water flow stability along the outside wall of the containment was studied. The results fit well with those correlations without airflow when the air velocity is less than 5.0 m/s. However, when the air velocity is larger than 5.0 m/s, the influence of the air velocity on the water film will appear and the mean water film thickness will be thicker. Based on the prototype operation parameters, experimental studies were carried and the results were compared with the Dittus-Boelter correlation within the operation ranges. A modification factor was proposed for the conservative application of this correlation for nuclear safety analysis.

Assessing verticalization effects on urban safety perception

OpenAIRE

Lourenço, Ricardo Barros

2017-01-01

We describe an experiment with the modeling of urban verticalization effects on perceived safety scores as obtained with computer vision on Google Streetview data for New York City. Preliminary results suggests that for smaller buildings (between one and seven floors), perceived safety increases with building height, but that for high-rise buildings, perceived safety decreases with increased height. We also determined that while height contributing for this relation, other zonal aspects also ...

Preliminary safety evaluation for the Forsmark area. Based on data and site descriptions after the initial site investigation stage

International Nuclear Information System (INIS)

Andersson, Johan

2005-08-01

The main objectives of this Preliminary Safety Evaluation (PSE) of the Forsmark area have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The evaluation shows that, even considering remaining uncertainties, the Forsmark area meets all stated safety requirements and preferences. Consequently, from a safety point of view, there is no reason not to continue the Site Investigations of the Forsmark area. There are still uncertainties to resolve and the safety would eventually need to be verified through a full safety assessment. Nevertheless, this Preliminary Safety Evaluation demonstrates that it is likely that a safe repository for spent nuclear fuel of the KBS-3 type could be constructed at the site. The following feedback is provided to the site investigations and the associated site modelling: Reducing the uncertainty on the deformation zone geometry inside the target area would be needed to more firmly define locations of the suitable deposition volumes. There is substantial uncertainty in the Discrete Fracture Network model. Further reduction of the uncertainties, if needed, would probably only be possible from the underground, detailed investigation phase. Efforts need also be spent on improving the DFN-modelling. There are assumptions made in current models that could be challenged and there seems to be room for better use of the borehole information. It is particularly important to provide

Preliminary safety evaluation for the Forsmark area. Based on data and site descriptions after the initial site investigation stage

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan [JA Streamflow AB, Aelvsjoe (Sweden)

2005-08-01

The main objectives of this Preliminary Safety Evaluation (PSE) of the Forsmark area have been to determine, with limited efforts, whether the feasibility study's judgement of the suitability of the candidate area with respect to long-term safety holds up in the light of the actual site investigation data; to provide feedback to continued site investigations and site-specific repository design and to identify site-specific scenarios and geoscientific issues for further analyses. The PSE focuses on comparing the attained knowledge of the sites with the suitability criteria as set out by SKB. The PSE does not aim at comparing sites and does not assess compliance with safety and radiation protection criteria. The evaluation shows that, even considering remaining uncertainties, the Forsmark area meets all stated safety requirements and preferences. Consequently, from a safety point of view, there is no reason not to continue the Site Investigations of the Forsmark area. There are still uncertainties to resolve and the safety would eventually need to be verified through a full safety assessment. Nevertheless, this Preliminary Safety Evaluation demonstrates that it is likely that a safe repository for spent nuclear fuel of the KBS-3 type could be constructed at the site. The following feedback is provided to the site investigations and the associated site modelling: Reducing the uncertainty on the deformation zone geometry inside the target area would be needed to more firmly define locations of the suitable deposition volumes. There is substantial uncertainty in the Discrete Fracture Network model. Further reduction of the uncertainties, if needed, would probably only be possible from the underground, detailed investigation phase. Efforts need also be spent on improving the DFN-modelling. There are assumptions made in current models that could be challenged and there seems to be room for better use of the borehole information. It is particularly important to

Preliminary safety analysis of high-intensity interval training (HIIT) in persons with chronic stroke.

Science.gov (United States)

Carl, Daniel L; Boyne, Pierce; Rockwell, Bradley; Gerson, Myron; Khoury, Jane; Kissela, Brett; Dunning, Kari

2017-03-01

The purpose of this study was to assess safety via electrocardiographic (ECG), blood pressure (BP), heart rate (HR), and orthopedic responses to 3 different high-intensity interval training (HIIT) protocols in persons with stroke. Eighteen participants (10 male; 61.9 + 8.3 years of age; 5.8 + 4.2 years poststroke) completed a symptom-limited graded exercise test (GXT) with ECG monitoring to screen for eligibility and determine HR peak. The 3 HIIT protocols involved repeated 30 s bursts of treadmill walking at maximum speed alternated with rest periods of 30 s (P30), 1 min (P60), or 2 min (P120). Sessions were performed in random order and included 5 min warm up, 20 min HIIT, and 5 min cool down. Variables measured included ECG activity, BP, HR, signs and symptoms of cardiovascular intolerance, and orthopedic concerns. Generalized linear mixed models and Tukey-Kramer adjustment were used to compare protocols using p HIIT session. HIIT elicited HRs in excess of 88% of measured HR peak including 6 (P30), 8 (P60), and 2 (P120) participants eliciting a HR response above their GXT HR peak . Both maximum BP and HR were significantly higher in P30 and P60 relative to P120. Preliminary data indicate that persons with chronic stroke who have been prescreened with an ECG stress test, a symptom-limited GXT, and a harness for fall protection may safely participate in HIIT, generating substantially higher HRs than what is seen in traditional moderate intensity training.

A measurement tool to assess culture change regarding patient safety in hospital obstetrical units.

Science.gov (United States)

Kenneth Milne, J; Bendaly, Nicole; Bendaly, Leslie; Worsley, Jill; FitzGerald, John; Nisker, Jeff

2010-06-01

Clinical error in acute care hospitals can only be addressed by developing a culture of safety. We sought to develop a cultural assessment survey (CAS) to assess patient safety culture change in obstetrical units. Interview prompts and a preliminary questionnaire were developed through a literature review of patient safety and "high reliability organizations," followed by interviews with members of the Managing Obstetrical Risk Efficiently (MOREOB) Program of the Society of Obstetricians and Gynaecologists of Canada. Three hundred preliminary questionnaires were mailed, and 21 interviews and 9 focus groups were conducted with the staff of 11 hospital sites participating in the program. To pilot test the CAS, 350 surveys were mailed to staff in participating hospitals, and interviews were conducted with seven nurses and five physicians who had completed the survey. Reliability analysis was conducted on four units that completed the CAS prior to and following the implementation of the first MOREOB module. Nineteen values and 105 behaviours, practices, and perceptions relating to patient safety were identified and included in the preliminary questionnaire, of which 143 of 300 (47.4%) were returned. Among the 220 cultural assessment surveys returned (62.9%), six cultural scales emerged: (1) patient safety as everyone's priority; (2) teamwork; (3) valuing individuals; (4) open communication; (5) learning; and (6) empowering individuals. The reliability analysis found all six scales to have internal reliability (Cronbach alpha), ranging from 0.72 (open communication) to 0.84 (valuing individuals). The CAS developed for this study may enable obstetrical units to assess change in patient safety culture.

Measuring Safety Culture on Ships Using Safety Climate: A Study among Indian Officers

Directory of Open Access Journals (Sweden)

Yogendra Bhattacharya

2015-12-01

Full Text Available Workplace safety continues to be an area of concern in the maritime industry due to the international nature of the operations. The effectiveness of extensive legislation to manage shipboard safety remains in doubt. The focus must therefore shift towards the human element - seafarers and their perceptions of safety. The study aims to understand the alignment that exists between safety culture and safety climate on board ships as perceived by seafarers. The underlying factors of safety climate were identified using factor analysis which isolated seven factors - Support on Safety, Organizational Support, Resource Availability, Work Environment, Job Demands, ‘Just’ Culture, and Safety Compliance. The perception of safety level of seafarers was found to be low indicating the existence of misalignments between safety culture values and the actual safety climate. The study also reveals that the safety perceptions of officers employed directly by ship owners and those by managers do not differ significantly, nor do they differ between senior and junior officers. A shift in perspective towards how seafarers themselves feel towards safety might provide more effective solutions – instead of relying on regulations - and indeed aid in reducing incidents on board. This paper details practical suggestions on how to identify the factors that contribute towards a better safety climate on board ships.

Safety design of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

1984-01-01

Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

Preliminary study of percutaneous nephrolithotomy on an ambulatory basis.

Science.gov (United States)

El-Tabey, Magdy Ahmed; Abd-Allah, Osama Abdel-Wahab; Ahmed, Ahmed Sebaey; El-Barky, Ehab Mohammed; Noureldin, Yasser Abdel-Sattar

2013-02-01

Preliminary study to assess the feasibility and safety of percutaneous nephrolithotomy (PCNL) as an ambulatory procedure. Between February 2011 and September 2012, 84 patients with renal calculi fulfilling the inclusion criteria were admitted to the Urology Department of Benha University Hospitals for PCNL. All patients were subjected to a full medical history, clinical, laboratory and radiological examinations. Tubeless PCNLs were done in the supine position, and an antegrade double-J stent was inserted. Operative time and intraoperative complications were recorded. Postoperatively, the hematocrit value, postoperative pain and analgesics, need of blood transfusion, stone-free rate, and length of hospital stay were recorded. Stable patients that could be safely discharged within 24 hours after surgery were considered ambulatory. All cases of tubeless PCNL were successfully done and no cases converted to open surgery. The overall stone-free rate was 91.7%, the mean postoperative pain score measured by the visual analog scale was 4.4 ± 1.2, the mean overall hematocrit deficit was 4.8 ± 2.2% and the mean hospital stay was 33.4 ± 17.5 hours. Ambulatory PCNL was accomplished in 60 out of 84 patients (71.4%) and double-J stents were removed 7-10 days postoperatively. In the non-ambulatory cases, double-J stents were removed after auxillary procedures were done according to each case. PCNL can be safely done on an ambulatory basis under strict criteria, but further studies are needed to confirm and expand these findings.

Seismic safety of Paks nuclear power plant

International Nuclear Information System (INIS)

Katona, T.

1995-01-01

This paper contains an overview of the results concerning the following activities: investigation of methods, regulations and techniques for reassessment of seismic safety of operating NPPs and upgrading of safety; investigation of earthquake hazards; development of concept for creation of the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept and preliminary evaluation of the seismic safety. It is limited on investigation of dynamic features of building structures, the building dynamical experiments and experimental investigation of the equipment

Preliminary simulation study of doppler reflectometry

International Nuclear Information System (INIS)

Ishii, Yuta; Hojo, Hitoshi; Yoshikawa, Masashi; Ichimura, Makoto; Haraguchi, Yusuke; Imai, Tsuyoshi; Mase, Atsushi

2010-01-01

A preliminary simulation study of Doppler reflectometry is performed. The simulations solve Maxwell's equations by a finite difference time domain (FDTD) code method in two dimensions. A moving corrugated metal target is used as a plasma cutoff layer to study the basic features of Doppler reflectometry. We examined the effects of the full width at half maximum (FWHM) of the electromagnetic waves and the corrugation depth of the metal target. Furthermore, the effect of a nonuniform plasma is studied using this FDTD analysis. The Doppler shift and velocity are compared with those obtained from FDTD analysis of a uniform plasma. (author)

Preliminary closed Brayton cycle study for a space reactor application

International Nuclear Information System (INIS)

Guimaraes, Lamartine Nogueira Frutuoso; Carvalho, Ricardo Pinto de; Camillo, Giannino Ponchio

2007-01-01

The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are to establish a starting concept for the CBCL components specifications, and to develop a demonstrative simulator of CBCL in nominal operation conditions. The ENU/IEAv preliminary design study is developing the CBCL around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO2 and gas mixtures such as helium and xenon. At this moment the simulator is running with Helium as the working fluid. Simplified models of heat and mass transfer are being developed to simulate thermal components. Future efforts will focus on keeping track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. (author)

Preliminary closed Brayton cycle study for a space reactor application

Energy Technology Data Exchange (ETDEWEB)

Guimaraes, Lamartine Nogueira Frutuoso; Carvalho, Ricardo Pinto de [Institute for Advanced Studies, Sao Jose dos Campos, SP (Brazil)]. E-mail: guimarae@ieav.cta.br; Camillo, Giannino Ponchio [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil)]. E-mail: gianninocamillo@gmail.com

2007-07-01

The Nuclear Energy Division (ENU) of the Institute for Advanced Studies (IEAv) has started a preliminary design study for a Closed Brayton Cycle Loop (CBCL) aimed at a space reactor application. The main objectives of the study are to establish a starting concept for the CBCL components specifications, and to develop a demonstrative simulator of CBCL in nominal operation conditions. The ENU/IEAv preliminary design study is developing the CBCL around the NOELLE 60290 turbo machine. The actual nuclear reactor study is being conducted independently. Because of that, a conventional heat source is being used for the CBCL, in this preliminary design phase. This paper describes the steady state simulator of the CBCL operating with NOELLE 60290 turbo machine. In principle, several gases are being considered as working fluid, as for instance: air, helium, nitrogen, CO2 and gas mixtures such as helium and xenon. At this moment the simulator is running with Helium as the working fluid. Simplified models of heat and mass transfer are being developed to simulate thermal components. Future efforts will focus on keeping track of the modifications being implemented at the NOELLE 60290 turbo machine in order to build the CBCL. (author)

Use of cross-linked carboxymethyl cellulose for soft-tissue augmentation: preliminary clinical studies

Directory of Open Access Journals (Sweden)

Mauro Leonardis

2010-11-01

Full Text Available Mauro Leonardis1, Andrea Palange2, Rodrigo FV Dornelles3, Felipe Hund41Department of Plastic Surgery, Salvator Mundi International Hospital, Roma, Italy; 2Department of Aesthetic Medicine, Fisiobios, Roma, Italy; 3Department of Plastic Surgery, Núcleo de Plástica Avançada, São Paulo, SP, Brazil; 4Department of Plastic Surgery, Consultorio de Cirurgia Plastica, Criciuma, SC, BrazilPurpose: The continual search for new products for soft-tissue augmentation has in recent years led to the introduction of long lasting alternatives to hyaluronic acids and collagen that are composed of other polymers able to improve clinical persistence over time. This is the first report in which sodium carboxymethyl cellulose (CMC has been chemically treated by the cross-linking process and thus used as a hydrogel for soft-tissue augmentation through injection with thin needles. The study evaluates, from a clinical point of view, the behavior of cross-linked carboxymethyl cellulose hydrogel used in the aesthetic field and its side effects so as to check the safety and performance of the polymer following intradermal injections.Patients and methods: This work shows the preliminary results of an ongoing clinical study conducted between 2006 and 2009, performed on 84 healthy volunteers (62 females, 22 males aged between 18 and 72 years, for the treatment of 168 nasolabial folds, 45 perioral wrinkles, and 39 lip volume.Results: Study results show an excellent correction of facial defects. Tolerance and aesthetic quality of the correction obtained indicate considerable safety features and absence of side effects. From a clinical point of view, hydrogel is gradually absorbed into the injection site without migration issues.Conclusion: Cross-linked CMC hydrogel proves to be an ideal agent for soft tissue augmentation with regard to safety and ease of application. It did not cause infection, extrusion, migration, or adverse reactions in the patients who have been

A preliminary safety evaluation of polyhexamethylene guanidine hydrochloride.

Science.gov (United States)

Asiedu-Gyekye, Isaac Julius; Mahmood, Seidu Abdulai; Awortwe, Charles; Nyarko, Alexander Kwadwo

2014-01-01

Polyhexamethylene guanidine hydrochloride (PHMGH) is used worldwide as an antimicrobial agent with broad spectra of activity and also for treating pool water. This non-GLP preliminary study aims at investigating in a subchronic toxicity study possible effects at supra-optimal doses of this biocide. Both acute and subchronic toxicity studies were conducted. LD(50) for PHMGH was estimated to be 600 mg/kg (ie LC(50) 2 ml of 7.5% solution) when administered as a single dose by gavage via a stomach tube in accordance with the expected route of administration. The acute studies showed that the median lethal dose (LD(50)) of 600 mg/kg was accompanied by signs of neurotoxicity. Haematological and biochemical parameters of subchronic toxicity studies were non-significant. Subchronic doses of 0.006 mg/kg, 0.012 mg/kg and 0.036 mg/kg were administered. 20% of the animals at a dose of 0.006 mg/kg and 0.036 mg/kg showed mild degrees of hydropic changes in proximal tubules while 10% of animals at all the doses had their liver tissues showing local areas of mild pericentral hepatocytes degeneration. PHMGH did not produce any major organ defect with regard to the kidney, heart, and liver. The LD(50) was much higher than the recommended dosage by a factor of about 50,000. The recommended residual concentration is far less than the median lethal dose using rats as test subjects. These results could serve as a basis for investigating the full toxicological profile if it is to be used for the treatment of raw water to make it potable. © The Author(s) 2014.

Development of the NUMO pre-selection, site-specific safety case

International Nuclear Information System (INIS)

Fujiyama, Tetsuo; Suzuki, Satoru; Deguchi, Akira; Umeki, Hiroyuki

2016-01-01

Key conclusions: ◆ “The NUMO pre-selection, site-specific safety case” provides the basic structure for subsequent safety cases that will be applied to any selected site, emphasising practical approaches and methodology which will be applicable for the conditions/constraints during an actual siting process. ◆ The preliminary results of the design and safety assessment would underpin the feasibility and safety of geological disposal in Japan.

Preliminary risk assessment of the Integral Inherently-Safe Light Water Reactor

International Nuclear Information System (INIS)

McCarroll, Kellen R.; Lee, John C.; Manera, Annalisa; Memmott, Matthew J.; Ferroni, Paolo

2017-01-01

The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) concept seeks to significantly increase nuclear power plant safety. The project implements a safety-by-design philosophy, eliminating several initiating events and providing novel, passive safety systems at the conceptual phase. Pursuit of unparalleled safety employs an integrated development process linking design with deterministic and probabilistic safety analyses. Unique aspects of the I 2 S-LWR concept and design process present challenges to the probabilistic risk assessment (PRA), particularly regarding overall flexibility, auditability and resolution of results. Useful approaches to initiating events and conditional failures are presented. To exemplify the risk-informed design process using PRA, a trade-off study of two safety system configurations is presented. Although further optimization is required, preliminary results indicate that the I 2 S-LWR can achieve a core damage frequency (CDF) from internal events less than 1.01 × 10 −8 /ry, including reactor vessel ruptures. Containment bypass frequency due to primary heat exchanger rupture is found to be comparable to non-vessel rupture CDF.

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

Preliminary design study of a steady state tokamak device

International Nuclear Information System (INIS)

Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

1992-09-01

Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

Joint SKI and SSI review of SKB preliminary safety assessment of repository for long-lived low- and intermediate-level waste. Review report

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

SKI and SSI find that SKB's first proper safety assessment of the SFL 3-5 repositories provides a valuable springboard for continued efforts in this field. Even though the safety assessment is relatively limited in scope, it has numerous merits. The specific problems associated with the chosen repository concept for SFL 3-5 are discussed in a generally transparent manner. On the other hand, the authorities consider that SKB have only partly achieved the expressed goal of studying the significance of the current repository design and the choice of site. The greatest deficiency consists in that neither internal disturbances (such as considerable cracking or degradation of concrete structures) nor external disturbances (such as the effects of climate changes and glaciation) have been addressed in a thorough manner. A coherent report justifying the design choice from a long-term safety perspective is, in large part, not found here. SKI and SSI recommend that SKB provide a comparison with other possible SFL 3-5 repository designs. Depending upon, among other factors, what geospheric and biospheric conditions are assumed, SKB have shown that the calculated dose values could be relatively high for certain cases. More realistic assessments would be needed to draw reasonable comparisons between different sites, and to evaluate the importance of different nuclides in different contexts. Our review of SKBs preliminary safety assessment indicates that a great deal of research and development work remains to be done before the level of knowledge in this field is comparable with that associated with the final repository for spent fuel. This is reflected with unanimity in the international expert committee's review, and in the consultants' reviews. SKI and SSI wish to point out in particular the fact that comparison with SFR is of limited value, since the safety associated with SFL 3- 5 must be assessed on a much longer time scale. SKI and SSI find it remarkable

Hydrothermal Liquefaction Treatment Preliminary Hazard Analysis Report

Energy Technology Data Exchange (ETDEWEB)

Lowry, Peter P.; Wagner, Katie A.

2015-08-31

A preliminary hazard assessment was completed during February 2015 to evaluate the conceptual design of the modular hydrothermal liquefaction treatment system. The hazard assessment was performed in 2 stages. An initial assessment utilizing Hazard Identification and Preliminary Hazards Analysis (PHA) techniques identified areas with significant or unique hazards (process safety-related hazards) that fall outside of the normal operating envelope of PNNL and warranted additional analysis. The subsequent assessment was based on a qualitative What-If analysis. This analysis was augmented, as necessary, by additional quantitative analysis for scenarios involving a release of hazardous material or energy with the potential for affecting the public.

Safety of High Speed Magnetic Levitation Transportation Systems: Preliminary Safety Review of the Transrapid Maglev System

Science.gov (United States)

1990-11-01

The safety of various magnetically levitated trains under development for possible : implementation in the United States is of direct concern to the Federal Railroad : Administration. This report, one in a series of planned reports on maglev safety, ...

Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

International Nuclear Information System (INIS)

Miller, W.J.

1980-05-01

This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan

Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

Energy Technology Data Exchange (ETDEWEB)

Miller, W.J.

1980-05-01

This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

Subseabed disposal safety analysis

International Nuclear Information System (INIS)

Koplick, C.M.; Kabele, T.J.

1982-01-01

This report summarizes the status of work performed by Analytic Sciences Corporation (TASC) in FY'81 on subseabed disposal safety analysis. Safety analysis for subseabed disposal is divided into two phases: pre-emplacement which includes all transportation, handling, and emplacement activities; and long-term (post-emplacement), which is concerned with the potential hazard after waste is safely emplaced. Details of TASC work in these two areas are provided in two technical reports. The work to date, while preliminary, supports the technical and environmental feasibility of subseabed disposal of HLW

A Preliminary Experimental Study on Flow Boiling CHF Characteristics of Ballooned Channel

International Nuclear Information System (INIS)

Kim, Yong Jin; Song, Sub Lee; Chang, Soon Heung; Moon, Sang Ki

2013-01-01

The purpose of this research is to measure heat transfer characteristics experimentally and to develop correlation based on experimental data. Experiments are in progress. The result of preliminary experimental test of ballooned channel was reported. The trends of CHF value for deformed channel is not usual as normal smooth tube. The spot of CHF was moved by changing different experimental cases. The transition of flow pattern at neck of deformation is considered as main factor of changing CHF trends. More cases are under operation and analysis based on flow dynamics are developing. Cladding is one of the most important parts in nuclear power plant because it is second barrier of radiation leakage from nuclear fuel. Originally, cladding keeps its integrity in 1200 .deg. C and 150bar, which is normal operation state of nuclear power plant. However, integrity of cladding can be deformed by more severe conditions caused by accident. In case of LOCA, high temperature, oxidation and thermal shock induced by safety injection can deform cladding. Main problem of deformed cladding is blockage of cooled to prevent core melt accident. Change of flow path by blockage affects flow of safety coolant, heat transfer coefficient and critical heat flux of rod bundles. Until now, there are insufficient heat transfer data for deformed flow path compared to normal flow path. In order to enhance safety of nuclear power plant after accident, it should be clarified that how deformed cladding affects heat transfer

Development of an X-ray installation for the study of secondary electrons: preliminary measurements and calculations

International Nuclear Information System (INIS)

Baguena, A.; Shaw, M.; Williart, A.; Baguena, A.; Garcia, G.

2006-01-01

We describe the calculations and preliminary measures made for the installation of a X-ray generator tube. This device is going to be used for the secondary electron production from photonic primary radiation of up to 125 keV. With this experimental system, we will study the energetic and space distribution of produced secondary electrons by obtaining its spectrum of energies and its angular distribution. This method of measurement is going to be applied in different targets of radiological, environmental and biological interest. Calculations in the present article include: theoretical yield of X-rays production of the designed equipment, necessary shielding for the radiological safety of the installation staff, and an estimated dose due to their use. Characteristics of the installation and the equipment are described with this purpose. (author)

Development of an X-ray installation for the study of secondary electrons: preliminary measurements and calculations

Energy Technology Data Exchange (ETDEWEB)

Baguena, A.; Shaw, M.; Williart, A. [Universidad Nacional de Educacion a Distancia, Dpto. Fisica de los Materiales, Madrid (Spain); Baguena, A. [Consejo de Seguridad Nuclear, Madrid (Spain); Garcia, G. [Instituto de Matematicas y Fisica Fundamental, Consejo Superior de Investigaciones Cientificas, Madrid (Spain)

2006-07-01

We describe the calculations and preliminary measures made for the installation of a X-ray generator tube. This device is going to be used for the secondary electron production from photonic primary radiation of up to 125 keV. With this experimental system, we will study the energetic and space distribution of produced secondary electrons by obtaining its spectrum of energies and its angular distribution. This method of measurement is going to be applied in different targets of radiological, environmental and biological interest. Calculations in the present article include: theoretical yield of X-rays production of the designed equipment, necessary shielding for the radiological safety of the installation staff, and an estimated dose due to their use. Characteristics of the installation and the equipment are described with this purpose. (author)

HTGR gas turbine power plant preliminary design

International Nuclear Information System (INIS)

Koutz, S.L.; Krase, J.M.; Meyer, L.

1973-01-01

The preliminary reference design of the HTGR gas turbine power plant is presented. Economic and practical problems and incentives related to the development and introduction of this type of power plant are evaluated. The plant features and major components are described, and a discussion of its performance, economics, development, safety, control, and maintenance is presented. 4 references

Railway safety climate: a study on organizational development.

Science.gov (United States)

Cheng, Yung-Hsiang

2017-09-07

The safety climate of an organization is considered a leading indicator of potential risk for railway organizations. This study adopts the perceptual measurement-individual attribute approach to investigate the safety climate of a railway organization. The railway safety climate attributes are evaluated from the perspective of railway system staff. We identify four safety climate dimensions from exploratory factor analysis, namely safety communication, safety training, safety management and subjectively evaluated safety performance. Analytical results indicate that the safety climate differs at vertical and horizontal organizational levels. This study contributes to the literature by providing empirical evidence of the multilevel safety climate in a railway organization, presents possible causes of the differences under various cultural contexts and differentiates between safety climate scales for diverse workgroups within the railway organization. This information can be used to improve the safety sustainability of railway organizations and to conduct safety supervisions for the government.

Fusion magnet safety studies program: superconducting magnet protection system and failure. Interim report

International Nuclear Information System (INIS)

Allinger, J.; Danby, G.; Hsieh, S.Y.; Keane, J.; Powell, J.; Prodell, A.

1975-11-01

This report includes the first two quarters study of available information on schemes for protecting superconducting magnets. These schemes can be divided into two different categories. The first category deals with the detection of faulty regions (or normal regions) in the magnet. The second category relates to the protection of the magnet when a fault is detected, and the derived signal which can be used to activate a safety system (or energy removal system). The general detection and protection methods are first described briefly and then followed by a survey of the protection systems used by different laboratories for various magnets. A survey of the cause of the magnet difficulties or failures is also included. A preliminary discussion of these protection schemes and the experimental development of this program is given

Preliminary 2D design study for A ampersand PCT

International Nuclear Information System (INIS)

Keto, E.; Azevedo, S.; Roberson, P.

1995-03-01

Lawrence Livermore National Laboratory is currently designing and constructing a tomographic scanner to obtain the most accurate possible assays of radioactivity in barrels of nuclear waste in a limited amount of time. This study demonstrates a method to explore different designs using laboratory experiments and numerical simulations. In particular, we examine the trade-off between spatial resolution and signal-to-noise. The simulations are conducted in two dimensions as a preliminary study for three dimensional imaging. We find that the optimal design is entirely dependent on the expected source sizes and activities. For nuclear waste barrels, preliminary results indicate that collimators with widths of 1 to 3 inch and aspect ratios of 5:1 to 10:1 should perform well. This type of study will be repeated in 3D in more detail to optimize the final design

Preliminary study of the specific endothelin a receptor antagonist zibotentan in combination with docetaxel in patients with metastatic castration-resistant prostate cancer.

Science.gov (United States)

Trump, Donald L; Payne, Heather; Miller, Kurt; de Bono, Johann S; Stephenson, Joe; Burris, Howard A; Nathan, Faith; Taboada, Maria; Morris, Thomas; Hubner, Andreas

2011-09-01

This two-part study assessed the safety and tolerability of combined treatment with zibotentan (ZD4054), a specific endothelin A receptor antagonist, plus docetaxel in patients with metastatic castration-resistant prostate cancer. Part A was an open-label, dose-finding phase to determine the safety and toxicity profile of zibotentan in combination with docetaxel. Patients received once-daily oral zibotentan 10 mg (initial cohort) or 15 mg in combination with docetaxel 75 mg/m(2) (administered on day 1 of each 21-day cycle) for up to 10 cycles. Part B was a double-blind phase which evaluated the safety and preliminary activity of zibotentan plus docetaxel. Patients were randomized 2:1 to receive zibotentan (at the highest tolerated dose identified in part A) plus docetaxel or placebo plus docetaxel. Six patients were enrolled in part A (n  = 3, zibotentan 10 mg; n = 3, zibotentan 15 mg). No dose-limiting toxicity was observed, thus zibotentan 15 mg in combination with docetaxel was evaluated in part B (n = 20, zibotentan plus docetaxel; n = 11, placebo plus docetaxel). CTCAE grade ≥3, most commonly neutropenia or leucopenia, were reported in 10 (50%) and nine (82%) patients in the zibotentan and placebo groups, respectively. One (17%) patient receiving placebo achieved complete response, two (22%) patients receiving zibotentan achieved partial response and stable disease occurred in six (67%) and three (50%) patients receiving zibotentan and placebo, respectively. The tolerability of zibotentan plus docetaxel was consistent with the known profiles of each drug. Sufficient preliminary activity was seen with this combination to merit continued development. Copyright © 2011 Wiley-Liss, Inc.

Study of the cost-benefit analysis method for safety. Meeting of the Permanent Group in charge of nuclear reactors on the 5 July 2007

International Nuclear Information System (INIS)

2007-07-01

After a recall of the history of the issue of third decennial visit of the 900 MW reactors, of the IRSN preliminary analysis, of elements given to the Permanent Group, of requests made by the ASN, and a presentation of the analysis performed by the IRSN, this large report presents the cost-benefit analysis method and its potential applications (principle, cost assessment, safety assessment, examples) and reports international experience gained in this area: the risk-informed approach (within the IAEA, in the USA, France and other European countries, the specific cost-benefit approach), existing cost-benefit type methods (comparison between methods used in the USA, in France and in Canada), and monetary assessment of accidents. It reports the application of the cost-benefit method for safety and its limitations, and then its application to modifications which have been implemented after safety re-examinations. It discusses the use of level 1 and 2 safety probabilistic studies, and reports the use of a cost-benefit method for safety within the frame of safety re-examinations

Fort Hood Solar Total Energy Project. Volume II. Preliminary design. Part 2. System performance and supporting studies. Final report

Energy Technology Data Exchange (ETDEWEB)

None,

1979-01-01

The preliminary design developed for the Solar Total Energy System to be installed at Fort Hood, Texas, is presented. System performance analysis and evaluation are described. Feedback of completed performance analyses on current system design and operating philosophy is discussed. The basic computer simulation techniques and assumptions are described and the resulting energy displacement analysis is presented. Supporting technical studies are presented. These include health and safety and reliability assessments; solar collector component evaluation; weather analysis; and a review of selected trade studies which address significant design alternatives. Additional supporting studies which are generally specific to the installation site are reported. These include solar availability analysis; energy load measurements; environmental impact assessment; life cycle cost and economic analysis; heat transfer fluid testing; meteorological/solar station planning; and information dissemination. (WHK)

IRSN preliminary considerations of the Fukushima event impact on the GENIV reactors

International Nuclear Information System (INIS)

Blanc, Daniel

2012-01-01

• The IRSN study aims to identify main specific safety issues for each GEN IV concept with regards to the European Nuclear Safety Regulatory Group (ENSREG) stress tests topics: → Earthquake; → Flooding; → Loss of the heat sink; →Loss of the power supply; → Combination of the two previous ones; → Severe accident management. • These main specific safety issues are identified as far as they could have a specific impact on: → Grace times; → Cliff edge effects; → Difficulties to cope with them. • The situation is different between existing reactors and for reactors not yet designed because the hazard level may be increase for the new reactors. • Nevertheless, the “hardened safety core” concept may be kept for extreme situations and will be identified on the basis of the above mentioned main specific safety issues. This analysis is a preliminary one based of the IRSN knowledge about the six GEN IV concepts issued from safety assessment already performed (in particular on the French SFRs already built) and publications

Nuclear fuels with high burnup: safety requirements

International Nuclear Information System (INIS)

Phuc Tran Dai

2016-01-01

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

Criteria for safety related nuclear plant operator actions: a preliminary assessment of available data

International Nuclear Information System (INIS)

Haas, P.M.; Bott, T.F.

1982-01-01

The need for a quantitative data base on the reliability of nuclear power plant operators has long been recognised by human factors and reliability analysts, and the great need for further assessment of operator performance under accident conditions has been dramatically emphasised by the incident at Three Mile Island-2. In the US, an effort has been under way for a number of years to develop a design standard to define when required manual operator action can be accepted as part of a nuclear plant design basis. To provide the necessary data base to support such standards and the necessary quantitative assessment of operator reliability, the US Nuclear Regulatory Commission is sponsoring a study at Oak Ridge National Laboratory to develop the data base. A preliminary assessment, completed in April 1979, concluded that sufficient data from US operating experience did not exist to provide an adequate data base. A programme of research using full-scope nuclear plant simulators and results that are correlated to field data was suggested. That programme was recently initiated. This paper reviews the approach, results and conclusions of the preliminary assessment and summarises the planned research programme of simulator studies. (author)

Regulatory Oversight of Safety Culture — Korea’s Experience

International Nuclear Information System (INIS)

Jung, S.J.; Choi, Y.S.; Kim, J.T.

2016-01-01

In Korea, a regulatory oversight program of safety culture was launched in 2012 to establish regulatory measures against several events caused by weak safety culture in the nuclear industry. This paper is intended to introduce the preliminary regulatory oversight framework, development and validation of safety culture components, pilot safety culture inspection results and lessons learned. The safety culture model should be based on a sound understanding of the national culture and industry characteristics where the model will be applied. The nuclear safety culture oversight model is being developed and built on the Korean regulatory system to independently assess the nuclear power operating organizations’ safety culture.

Safety studies dedicated to molten salt reactors with a fast neutron spectrum and operated in the Thorium fuel cycle - Innovative concept of Molten Salt Fast Reactor

International Nuclear Information System (INIS)

Brovchenko, Mariya

2013-01-01

The nuclear reactors of the 4. generation must allow an optimized use of natural resources, while performing at a high safety level. The framework of this thesis is the deployment study of one of such a system, an innovative and still little studied Molten Salt Fast Reactor. An excellent safety is an ultimate requirement of the nuclear energy deployment, so it is important to raise this question at the current early stage of the MSFR concept development. This concept was the subject of a neutronic tool benchmark within a European project EVOL. Definition, calculations and results analyses were performed during this thesis. Comparisons of static neutronic and burn-up calculations, performed by the project participants, concluded to a good agreement between the different codes and methods used and pointed out the sensibility of the nuclear database choice on the results. With the aim of safety analysis of the MSFR, the decay heat was studied in detail. The tool used for the decay heat calculation was developed and validated, to finally evaluate the decay heat in the reactor. The decay heat source presented in different zones was quantified, concluding to a high importance of the cooling of the fuel salt and the bubbling system enclosing a part of the fission products. The safety analysis methodology was also studied in this thesis. Even if the safety principles are directly transposable to the MSFR, the precise recommendations are not. This is due to the specificity of the design that relies on the liquid state of the fuel, on the reprocessing systems located in the reactor and the embryonic stage of the design. First, a preliminary transposition work of some criteria to the MSFR design was realized, resulting amongst other things in a list of accidental scenarios particular for MSFR. Finally, a preliminary physical study of some types of accidental scenarios was performed, that can be used as a basis for further analyses with more sophisticated tools. (author) [fr

Solid waste burial grounds interim safety analysis

International Nuclear Information System (INIS)

Saito, G.H.

1994-01-01

This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

Solid waste burial grounds interim safety analysis

Energy Technology Data Exchange (ETDEWEB)

Saito, G.H.

1994-10-01

This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

Operation safety of control systems. Principles and methods

International Nuclear Information System (INIS)

Aubry, J.F.; Chatelet, E.

2008-01-01

This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)

A Reliability Assessment Method for the VHTR Safety Systems

International Nuclear Information System (INIS)

Lee, Hyung Sok; Jae, Moo Sung; Kim, Yong Wan

2011-01-01

The Passive safety system by very high temperature reactor which has attracted worldwide attention in the last century is the reliability safety system introduced for the improvement in the safety of the next generation nuclear power plant design. The Passive system functionality does not rely on an external source of energy, but on an intelligent use of the natural phenomena, such as gravity, conduction and radiation, which are always present. Because of these features, it is difficult to evaluate the passive safety on the risk analysis methodology having considered the existing active system failure. Therefore new reliability methodology has to be considered. In this study, the preliminary evaluation and conceptualization are tried, applying the concept of the load and capacity from the reliability physics model, designing the new passive system analysis methodology, and the trial applying to paper plant.

Second preliminary design of JAERI experimental fusion reactor (JXFR)

International Nuclear Information System (INIS)

Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

1979-06-01

Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

Measuring Health Information Dissemination and Identifying Target Interest Communities on Twitter: Methods Development and Case Study of the @SafetyMD Network.

Science.gov (United States)

Kandadai, Venk; Yang, Haodong; Jiang, Ling; Yang, Christopher C; Fleisher, Linda; Winston, Flaura Koplin

2016-05-05

Little is known about the ability of individual stakeholder groups to achieve health information dissemination goals through Twitter. This study aimed to develop and apply methods for the systematic evaluation and optimization of health information dissemination by stakeholders through Twitter. Tweet content from 1790 followers of @SafetyMD (July-November 2012) was examined. User emphasis, a new indicator of Twitter information dissemination, was defined and applied to retweets across two levels of retweeters originating from @SafetyMD. User interest clusters were identified based on principal component analysis (PCA) and hierarchical cluster analysis (HCA) of a random sample of 170 followers. User emphasis of keywords remained across levels but decreased by 9.5 percentage points. PCA and HCA identified 12 statistically unique clusters of followers within the @SafetyMD Twitter network. This study is one of the first to develop methods for use by stakeholders to evaluate and optimize their use of Twitter to disseminate health information. Our new methods provide preliminary evidence that individual stakeholders can evaluate the effectiveness of health information dissemination and create content-specific clusters for more specific targeted messaging.

Geoscientific long-term prognosis. Preliminary safety analysis for the site Gorleben; Geowissenschaftliche Langzeitprognose. Bericht zum Arbeitspaket 2. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

Energy Technology Data Exchange (ETDEWEB)

Mrugalla, Sabine [Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover (Germany)

2011-07-15

The preliminary safety analysis of the site Gorleben includes the following chapters: (1) Introduction; (2) Aim and content of the geoscientific long-term prognosis for the site Gorleben; (3) Boundary conditions at the site Gorleben: climate; geomorphology; overlying rocks and adjoining rocks; hydrogeology; salt deposit Gorleben. (4) Probable future geological developments at the site Gorleben: supraregional developments with effects on the site Gorleben; glacial period developments; developments of the geomorphology, overlying and adjoining rocks; future developments of the hydrological systems at the site Gorleben; future saliniferous specific developments of the salt deposit Gorleben. (5) Commentary on the unlikely or excludable developments of the site Gorleben.

Dataset for Phase I randomized clinical trial for safety and tolerability of GET 73 in single and repeated ascending doses including preliminary pharmacokinetic parameters

Directory of Open Access Journals (Sweden)

Carolina L. Haass-Koffler

2017-12-01

Full Text Available The data in this article outline the methods used for the administration of GET 73 in the first time-in-human manuscript entitled “Phase I randomized clinical trial for the safety, tolerability and preliminary pharmacokinetics of the mGluR5 negative allosteric modulator GET 73 following single and repeated doses in healthy male volunteers” (Haass-Koffler et al., 2017 [1]. Data sets are provided in two different manners. The first series of tables provided includes procedural information about the experiments conducted. The next series of tables provided includes Pharmacokinetic (PK parameters for GET 73 and its main metabolite MET 2. This set of data is comprised by two experiments: Experiment 1 references a single ascending dose administration of GET 73 and Experiment 2 references a repeated ascending dose administration of GET 73. Keywords: Glutamate receptor subtype 5 (mGlu5, Allosteric modulator, GET 73, Safety, Tolerability

A guide to introducing burnup credit, preliminary version (English translation)

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

2017-06-01

There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)

Endobronchial Occlusion Stent: A Preliminary Experimental Study

Energy Technology Data Exchange (ETDEWEB)

Choi, Yo Won; Jeong, Woo Kyoung; Lee, Seoung Hoon; Heo, Jeong Nam; Jeon, Seok Chol [Hanyang University College of Medicine, Seoul (Korea, Republic of); Ko, Gi Young; Song, Ho Young [University of Ulsan College of Medicine, Ulsan (Korea, Republic of)

2010-04-15

To evaluate the safety and the technical feasibility of the use of an endobronchial occlusion stent and to get preliminary data for the development of the optimal material required for endobronchial occlusions. A commercialized, self-expandable tracheobronchial stent was modified; one half had a polyurethane cover with an occluded end and the other half was uncovered with a flaring configuration. The occluded end was placed such that it would face the distal lung. Under fluoroscopic guidance, seven stents were placed at the lower lobar bronchus in 6 mini-pigs. The bronchial obstruction was examined immediately after stent placement. Chest radiographs were taken at days 1, 7, 14, and 28 after stent placement and the removed airways from two, two, one, and one mini-pigs sacrificed on corresponding days were examined for the maintenance of bronchial obstruction. Stents were successfully placed and induced the immediate bronchial obstruction in all mini-pigs. Five of seven airways with occlusion stents maintained an obstruction until the mini-pigs were sacrificed. Proximal stent migration occurred in two mini-pigs (29%), and pulmonary consolidations were observed distal to four of the stents (57%). The placement of an endobronchial occlusion stent and the obstruction of targeted bronchi seem to be feasible, but an add-on check valve should be considered to prevent stent migration and obstructive pneumonia

Selected problems and results of the transient event and reliability analyses for the German safety study

International Nuclear Information System (INIS)

Hoertner, H.

1977-01-01

For the investigation of the risk of nuclear power plants loss-of-coolant accidents and transients have to be analyzed. The different functions of the engineered safety features installed to cope with transients are explained. The event tree analysis is carried out for the important transient 'loss of normal onsite power'. Preliminary results of the reliability analyses performed for quantitative evaluation of this event tree are shown. (orig.) [de

Defining safety culture and the nexus between safety goals and safety culture. 1. An Investigation Study on Practical Points of Safety Management

International Nuclear Information System (INIS)

Hasegawa, Naoko; Takano, Kenichi; Hirose, Ayako

2001-01-01

In a report after the Chernobyl accident, the International Atomic Energy Agency indicated the definition and the importance of safety culture and the ideal organizational state where safety culture pervades. However, the report did not mention practical approaches to enhance safety culture. In Japan, although there had been investigations that clarified the consciousness of employees and the organizational climate in the nuclear power and railway industries, organizational factors that clarified the level of organization safety and practical methods that spread safety culture in an organization had not been studied. The Central Research Institute of the Electric Power Industry conducted surveys of organizational culture for the construction, chemical, and manufacturing industries. The aim of our study was to clarify the organizational factors that influence safety in an organization expressed in employee safety consciousness, commitment to safety activities, rate of accidents, etc. If these areas were clarified, the level of organization safety might be evaluated, and practical ways could be suggested to enhance the safety culture. Consequently, a series of investigations was conducted to clarify relationships among organizational climate, employee consciousness, safety management and activities, and rate of accidents. The questionnaire surveys were conducted in 1998-1999. The subjects were (a) managers of the safety management sections in the head offices of the construction, chemical, and manufacturing industries; (b) responsible persons in factories of the chemical and manufacturing industries; and (c) general workers in factories of the chemical and manufacturing industries. The number of collected data was (a) managers in the head office: 48 from the construction industry and 58 from the chemical and manufacturing industries, (b) responsible persons in factories: 567, and (c) general workers: from 29 factories. Items in the questionnaires were selected from

Applying interprofessional Team-Based Learning in patient safety: a pilot evaluation study.

Science.gov (United States)

Lochner, Lukas; Girardi, Sandra; Pavcovich, Alessandra; Meier, Horand; Mantovan, Franco; Ausserhofer, Dietmar

2018-03-27

.' Findings on safety attitudes and behaviours were mixed. TBL was well received by the students. Our first findings indicate that interprofessional TBL seems to be a promising pedagogical method to achieve patient safety learning objectives. It is crucial to develop relevant clinical cases that involve all professions. Further research with larger sample sizes (e.g. including medical students) and more rigorous study designs (e.g. pre-test post-test with a control group) is needed to confirm our preliminary findings.

To dimension safety valves. Probabilist study

International Nuclear Information System (INIS)

Noel, Robert; Couvreur, Denis

1982-01-01

The gauge of safety valves of a steam pressure apparatus is usually determined according to an operating situation envelope which it is admitted covers all that can happen in reality. For the safety of the dryer-superheaters of turbines in nuclear power stations, Electricite de France and Alsthom-Atlantique made a reliability study; its method is exposed and the results are discussed. Such a study is heavy going and complex, but in return it permits a better quantitative understanding of the various dimension and operating parameters of an installation which condition its safety. It is therefore a source of progress [fr

Preliminary ATWS analysis for the IRIS PRA

International Nuclear Information System (INIS)

Maddalena Barra; Marco S Ghisu; David J Finnicum; Luca Oriani

2005-01-01

Full text of publication follows: The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002. IRIS has been primarily focused on establishing a design with innovative safety characteristics. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. In IRIS, this concept is implemented through the 'safety by design' approach, which allows to minimize the number and complexity of the safety systems and required operator actions. The end result is a design with significantly reduced complexity and improved operability, and extensive plant simplifications to enhance construction. To support the optimization of the plant design and confirm the effectiveness of the safety by design approach in mitigating or eliminating events and thus providing a significant reduction in the probability of severe accidents, the PRA is being used as an integral part of the design process. A preliminary but extensive Level 1 PRA model has been developed to support the pre-application licensing of the IRIS design. As a result of the Preliminary IRIS PRA, an optimization of the design from a reliability point of view was completed, and an extremely low (about 1.2 E -8 ) core damage frequency (CDF) was assessed to confirm the impact of the safety by design approach. This first assessment is a result of a PRA model including internal initiating events. During this assessment, several assumptions were necessary to complete the CDF evaluation. In particular Anticipated Transients Without Scram (ATWS) were not included in this initial assessment, because their contribution to core damage frequency was assumed

GRIST-2 preliminary test plan and requirements for fuel fabrication and preirradiation

International Nuclear Information System (INIS)

Tang, I.M.; Harmon, D.P.; Torri, A.

1978-12-01

The preliminary version of the GRIST-2 test plan has been developed for the planned initial 5 years (1984 to 1989) of TREAT-Upgrade in-pile tests. These tests will be employed to study the phenomenology and integral behavior of GCFR core disruptive accidents (CDAs) and to support the Final Safety Analysis Report (FSAR) CDA analyses for the demonstration plant licensing. The preliminary test plan is outlined. Test Phases I and II are for the fresh fuel (preconditioned or not) CDA behavior at the beginning-of-life (BOL) reactor state. Phase III is for the reactor state that contains irradiated fuel with a saturated content of helium and fission gas. Phase IV is for larger bundle tests and scaling effects

Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Katanishi, Shoji; Kunitomi, Kazuhiko

2003-01-01

Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. (author)

Model quality and safety studies

DEFF Research Database (Denmark)

Petersen, K.E.

1997-01-01

The paper describes the EC initiative on model quality assessment and emphasizes some of the problems encountered in the selection of data from field tests used in the evaluation process. Further, it discusses the impact of model uncertainties in safety studies of industrial plants. The model...... that most of these have never been through a procedure of evaluation, but nonetheless are used to assist in making decisions that may directly affect the safety of the public and the environment. As a major funder of European research on major industrial hazards, DGXII is conscious of the importance......-tain model is appropriate for use in solving a given problem. Further, the findings from the REDIPHEM project related to dense gas dispersion will be highlighted. Finally, the paper will discuss the need for model quality assessment in safety studies....

In-pile experimental facility needs for LMFR safety research

International Nuclear Information System (INIS)

Kawata, Norio; Niwa, Hajime

1994-01-01

Although the achievement of the safety research during the past years has been significant, there still exists a strong need for future research, especially when there is prospect for future LMFR commercialization. In this paper, our current views are described on future research needs especially with a new in-pile experimental facility. The basic ideas and progress are outlined of a preliminary feasibility study. (author)

Safety implications of control systems

International Nuclear Information System (INIS)

Smith, O.L.

1983-01-01

The Safety Implications of Control Systems Program has three major activities in support of USI-A47. The first task is a failure mode and effects analysis of all plant systems which may potentially induce control system disturbance that have safety implications. This task has made a preliminary study of overfill events and recommended cases for further analysis on the hybrid simulator. Work continues on overcooling and undercooling. A detailed investigation of electric power network is in progress. LERs are providing guidance on important failure modes that will provide initial conditions for further simulator studies. The simulator taks is generating a detailed model of the control system supported by appropriate neutronics, hydraulics, and thermodynamics submodels of all other principal plant components. The simulator is in the last stages of development. Checkout calculations are in progress to establish model stability, robustness, and qualitative credibility. Verification against benchmark codes and plant data will follow

Muon-catalyzed fusion experiment target and detector system. Preliminary design report

International Nuclear Information System (INIS)

Jones, S.E.; Watts, K.D.; Caffrey, A.J.; Walter, J.B.

1982-03-01

We present detailed plans for the target and particle detector systems for the muon-catalyzed fusion experiment. Requirements imposed on the target vessel by experimental conditions and safety considerations are delineated. Preliminary designs for the target vessel capsule and secondary containment vessel have been developed which meet these requirements. In addition, the particle detection system is outlined, including associated fast electronics and on-line data acquisition. Computer programs developed to study the target and detector system designs are described

Preliminary Safety Analysis of the Gorleben Site: Safety Concept and Application to Scenario Development Based on a Site-Specific Features, Events and Processes (FEP) Database - 13304

Energy Technology Data Exchange (ETDEWEB)

Moenig, Joerg; Beuth, Thomas; Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Theodor-Heuss-Str. 4, D-38122 Braunschweig (Germany); Lommerzheim, Andre [DBE TECHNOLOGY GmbH, Eschenstr. 55, D-31224 Peine (Germany); Mrugalla, Sabine [Federal Institute for Geosciences and Natural Resources (BGR), Stilleweg 2, D-30655 Hannover (Germany)

2013-07-01

Based upon the German safety criteria, released in 2010 by the Federal Ministry of the Environment (BMU), a safety concept and a safety assessment concept for the disposal of heat-generating high-level waste have both been developed in the framework of the preliminary safety case for the Gorleben site (Project VSG). The main objective of the disposal is to contain the radioactive waste inside a defined rock zone, which is called containment-providing rock zone. The radionuclides shall remain essentially at the emplacement site, and at the most, a small defined quantity of material shall be able to leave this rock zone. This shall be accomplished by the geological barrier and a technical barrier system, which is required to seal the inevitable penetration of the geological barrier by the construction of the mine. The safe containment has to be demonstrated for probable and less probable evolutions of the site, while evolutions with very low probability (less than 1 % over the demonstration period of 1 million years) need not to be considered. Owing to the uncertainty in predicting the real evolution of the site, plausible scenarios have been derived in a systematic manner. Therefore, a comprehensive site-specific features, events and processes (FEP) data base for the Gorleben site has been developed. The safety concept was directly taken into account, e.g. by identification of FEP with direct influence on the barriers that provide the containment. No effort was spared to identify the interactions of the FEP, their probabilities of occurrence, and their characteristics (values). The information stored in the data base provided the basis for the development of scenarios. The scenario development methodology is based on FEP related to an impairment of the functionality of a subset of barriers, called initial barriers. By taking these FEP into account in their probable characteristics the reference scenario is derived. Thus, the reference scenario describes a

Preliminary summary of the ETF conceptual studies

Science.gov (United States)

Seikel, G. R.; Bercaw, R. W.; Pearson, C. V.; Owens, W. R.

1978-01-01

Power plant studies have shown the attractiveness of MHD topped steam power plants for baseload utility applications. To realize these advantages, a three-phase development program was initiated. In the first phase, the engineering data and experience were developed for the design and construction of a pilot plant, the Engineering Test Facility (ETF). Results of the ETF studies are reviewed. These three parallel independent studies were conducted by industrial teams led by the AVCO Everett Research Laboratory, the General Electric Corporation, and the Westinghouse Corporation. A preliminary analysis and the status of the critical evaluation of these results are presented.

A preliminary neutron crystallographic study of thaumatin

Energy Technology Data Exchange (ETDEWEB)

Teixeira, Susana C. M. [ILL-EMBL Deuteration Laboratory, Partnership for Structural Biology, 6 Rue Jules Horowitz, 38042 Grenoble (France); Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom); Blakeley, Matthew P. [Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); Leal, Ricardo M. F. [ILL-EMBL Deuteration Laboratory, Partnership for Structural Biology, 6 Rue Jules Horowitz, 38042 Grenoble (France); Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom); ESRF, 6 Rue Jules Horowitz, BP-220, 38043 Grenoble (France); Mitchell, Edward P. [EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom); ESRF, 6 Rue Jules Horowitz, BP-220, 38043 Grenoble (France); Forsyth, V. Trevor, E-mail: tforsyth@ill.fr [ILL-EMBL Deuteration Laboratory, Partnership for Structural Biology, 6 Rue Jules Horowitz, 38042 Grenoble (France); Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France); EPSAM and ISTM, Keele University, Staffordshire ST5 5BG (United Kingdom)

2008-05-01

Preliminary neutron crystallographic data from the sweet protein thaumatin have been recorded using the LADI-III diffractometer at the Institut Laue Langevin (ILL). The results illustrate the feasibility of a full neutron structural analysis aimed at further understanding the molecular basis of the perception of sweet taste. Such an analysis will exploit the use of perdeuterated thaumatin. A preliminary neutron crystallographic study of the sweet protein thaumatin is presented. Large hydrogenated crystals were prepared in deuterated crystallization buffer using the gel-acupuncture method. Data were collected to a resolution of 2 Å on the LADI-III diffractometer at the Institut Laue Langevin (ILL). The results demonstrate the feasibility of a full neutron crystallographic analysis of this structure aimed at providing relevant information on the location of H atoms, the distribution of charge on the protein surface and localized water in the structure. This information will be of interest for understanding the specificity of thaumatin–receptor interactions and will contribute to further understanding of the molecular mechanisms underlying the perception of taste.

Comparative studies of CERCER and CERMET fuels for EFIT from the viewpoint of core performance and safety

International Nuclear Information System (INIS)

Chen, X.N.; Rineiski, A.; Maschek, W.; Liu, P.; Boccaccini, C.M.; Sobolev, V.; Delage, F.; Rimpault, G.

2011-01-01

The European Facility for Industrial Transmutation (EFIT) has been developed within the 6. EU Framework by the EUROTRANS Program, aiming at a generic conceptual design of an accelerator driven transmuter. This paper deals with assessments of EFIT cores with CERCER and CERMET fuels from the viewpoint of core performance and safety. The conclusive remarks can be drawn as follows. Because of its much better thermal conductivity, the CERMET core can be designed by using thicker pins, so that it has the same or even better transmutation performance compared to the CERCER core. Both CERCER and CERMET fuels fulfill safety requirements. Moreover the CERMET fuel has higher fuel safety margins than the CERCER one. Preliminary analyses show that the CERMET total core power can be further increased by 50% at least without exceeding fuel and clad temperature limits. (authors)

MALLARD REPRODUCTIVE TESTING IN A POND ENVIRONMENT: A PRELIMINARY STUDY

Science.gov (United States)

A 2-year preliminary study was conducted on mallard ducks to determine the feasibility of using outdoor pond enclosures for reproductive studies and to evaluate the effects of the insecticide chlorpyrifos on mallard reproduction. No significant reproductive effects were observed ...

Preliminary waste acceptance requirements - Konrad repository project

International Nuclear Information System (INIS)

Brennecke, P.W.; Warnecke, E.H.

1991-01-01

In Germany, the planned Konrad repository is proposed for the disposal of all types of radioactive wastes whose thermal influence upon the host rock is negligible. The Bundesamt fuer Strahlenschutz has established Preliminary Waste Acceptance Requirements (as of April 1990) for this facility. The respective requirements were developed on the basis of the results of site-specific safety assessments. They include general requirements on the waste packages to be disposed of as well as more specific requirements on the waste forms, the packaging and the radionuclide inventory per waste package. In addition, the delivery of waste packages was regulated. An outline of the structure and the elements of the Preliminary Waste Acceptance Requirements of April 1990 is given including comments on their legal status. (Author)

77 FR 27776 - Safety and Occupational Health Study Section (SOHSS), National Institute for Occupational Safety...

Science.gov (United States)

2012-05-11

... Occupational Health Study Section (SOHSS), National Institute for Occupational Safety and Health (NIOSH) In... Services Office, CDC, pursuant to Public Law 92-463. Purpose: The Safety and Occupational Health Study... standard grants review and funding cycles pertaining to research issues in occupational safety and health...

76 FR 18220 - Safety and Occupational Health Study Section (SOHSS), National Institute for Occupational Safety...

Science.gov (United States)

2011-04-01

... Occupational Health Study Section (SOHSS), National Institute for Occupational Safety and Health (NIOSH) In... Services Office, CDC, pursuant to Public Law 92-463. Purpose: The Safety and Occupational Health Study... standard grants review and funding cycles pertaining to research issues in occupational safety and health...

A New Silver Complex with Ofloxacin – Preliminary Study

Directory of Open Access Journals (Sweden)

Rusu Aura

2016-06-01

Full Text Available Objective: Silver complexes of antibacterial quinolones have the potential advantage of combining the antibacterial activity of silver and fluoroquinolones. The objective of our study was the preparation and the preliminary physico-chemical characterization of a silver complex with ofloxacin.

Human Resources Readiness as TSO for Deterministic Safety Analysis on the First NPP in Indonesia

International Nuclear Information System (INIS)

Sony Tjahyani, D. T.

2010-01-01

In government regulation no. 43 year 2006 it is mentioned that preliminary safety analysis report and final safety analysis report are one of requirements which should be applied in construction and operation licensing for commercial power reactor (NPPs). The purpose of safety analysis report is to confirm the adequacy and efficiency of provisions within the defence in depth of nuclear reactor. Deterministic analysis is used on the safety analysis report. One of the TSO task is to evaluate this report based on request of operator or regulatory body. This paper discusses about human resources readiness as TSO for deterministic safety analysis on the first NPP in Indonesia. The assessment is done by comparing the analysis step on SS-23 and SS-30 with human resources status of BATAN currently. The assessment results showed that human resources for deterministic safety analysis are ready as TSO especially to review preliminary safety analysis report and to revise final safety analysis report in licensing on the first NPP in Indonesia. Otherwise, to prepare the safety analysis report is still needed many competency human resources. (author)

A study for structural safety of ISER reactor building under impact load

International Nuclear Information System (INIS)

Takeuchi, Yoichiro; Hasegawa, Toshiyasu; Mutoh, Atsushi; Wakabayashi, Hiroaki.

1991-01-01

ISER (Inherently Safe and Economical Reactor) proposed in Japan by an academic circle and industries is expected to be used world-wide particularly in developing countries where an energy crunch is feared in the 21-st century. A certain level of hardened structures for plant safety seems to be effective and may be required by the regulatory body, since the ISER is claimed to be inherently safe even against a kind of external load. This paper concerns impact resistant design of ISER. A brief state-of-the-art review on related works, impact resistant design flow and results of some preliminary analysis of a proposed ISER model is also presented. (author)

Fusion reactor safety studies, FY 1977

International Nuclear Information System (INIS)

Darby, J.B. Jr.

1978-04-01

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

A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

75 FR 26266 - Safety and Occupational Health Study Section (SOHSS), National Institute for Occupational Safety...

Science.gov (United States)

2010-05-11

... Occupational Health Study Section (SOHSS), National Institute for Occupational Safety and Health (NIOSH) In...) Public Law 92-463. Purpose: The Safety and Occupational Health Study Section will review, discuss, and... cycles pertaining to research issues in occupational safety and health, and allied areas. It is the...

Repository Subsurface Preliminary Fire Hazard Analysis

International Nuclear Information System (INIS)

Logan, Richard C.

2001-01-01

This fire hazard analysis identifies preliminary design and operations features, fire, and explosion hazards, and provides a reasonable basis to establish the design requirements of fire protection systems during development and emplacement phases of the subsurface repository. This document follows the Technical Work Plan (TWP) (CRWMS M and O 2001c) which was prepared in accordance with AP-2.21Q, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities''; Attachment 4 of AP-ESH-008, ''Hazards Analysis System''; and AP-3.11Q, ''Technical Reports''. The objective of this report is to establish the requirements that provide for facility nuclear safety and a proper level of personnel safety and property protection from the effects of fire and the adverse effects of fire-extinguishing agents

Development and preliminary experimental study on micro-stacked insulator

International Nuclear Information System (INIS)

Ren Chengyan; Yuan Weiqun; Zhang Dongdong; Yan Ping; Wang Jue

2009-01-01

High gradient insulating technology is one of the key technologies in new type dielectric wall accelerator(DWA). High gradient insulator, namely micro-stacked insulator, was developed and preliminary experimental study was done. Based on the finite element and particle simulating method, surface electric field distribution and electron movement track of micro-stacked insulator were numerated, and then the optimized design proposal was put forward. Using high temperature laminated method, we developed micro-stacked insulator samples which uses exhaustive fluorinated ethylene propylene(FEP) as dielectric layer and stainless steel as metal layer. Preliminary experiment of vacuum surface flashover in nanosecond pulse voltage was done and micro-stacked insulator exhibited favorable vacuum surface flashover performance with flashover field strength of near 180 kV/cm. (authors)

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

Science.gov (United States)

Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

2012-07-01

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

Applying principles from safety science to improve child protection.

Science.gov (United States)

Cull, Michael J; Rzepnicki, Tina L; O'Day, Kathryn; Epstein, Richard A

2013-01-01

Child Protective Services Agencies (CPSAs) share many characteristics with other organizations operating in high-risk, high-profile industries. Over the past 50 years, industries as diverse as aviation, nuclear power, and healthcare have applied principles from safety science to improve practice. The current paper describes the rationale, characteristics, and challenges of applying concepts from the safety culture literature to CPSAs. Preliminary efforts to apply key principles aimed at improving child safety and well-being in two states are also presented.

KALIMER preliminary conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

2000-08-01

This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report.

KALIMER preliminary conceptual design report

International Nuclear Information System (INIS)

Hahn, Do Hee; Kim, Y. J.; Kim, Y. G. and others

2000-08-01

This report, which summarizes the result of preliminary conceptual design activities during Phase 1, follows the format of safety analysis report. The purpose of publishing this report is to gather all of the design information developed so far in a systematic way so that KALIMER designers have a common source of the consistent design information necessary for their future design activities. This report will be revised and updated as design changes occur and more detailed design specification is developed during Phase 2. Chapter 1 describes the KALIMER Project. Chapter 2 includes the top level design requirements of KALIMER and general plant description. Chapter 3 summarizes the design of structures, components, equipment and systems. Specific systems and safety analysis results are described in the remaining chapters. Appendix on the HCDA evaluation is attached at the end of this report

Oxygenates in automotive fuels. Consequence analysis - preliminary study

International Nuclear Information System (INIS)

Brandberg, Aa.; Saevbark, B.

1994-01-01

Oxygenates is used in gasoline due to several reasons. They are added as high-octane components in unleaded gasoline and as agents to reduce the emission of harmful substances. Oxygenates produced from biomass might constitute a coming market for alternative fuels. This preliminary study describes the prerequisites and consequences of such an oxygenate utilization. 39 refs, 9 figs, 5 tabs

Safety analysis and risk assessment of the National Ignition Facility

International Nuclear Information System (INIS)

Brereton, S.; McLouth, L.; Odell, B.

1996-01-01

The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF and the methodology used to study them. It provides a summary of the methodology, an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low

Workforce perceptions of hospital safety culture: development and validation of the patient safety climate in healthcare organizations survey.

Science.gov (United States)

Singer, Sara; Meterko, Mark; Baker, Laurence; Gaba, David; Falwell, Alyson; Rosen, Amy

2007-10-01

To describe the development of an instrument for assessing workforce perceptions of hospital safety culture and to assess its reliability and validity. Primary data collected between March 2004 and May 2005. Personnel from 105 U.S. hospitals completed a 38-item paper and pencil survey. We received 21,496 completed questionnaires, representing a 51 percent response rate. Based on review of existing safety climate surveys, we developed a list of key topics pertinent to maintaining a culture of safety in high-reliability organizations. We developed a draft questionnaire to address these topics and pilot tested it in four preliminary studies of hospital personnel. We modified the questionnaire based on experience and respondent feedback, and distributed the revised version to 42,249 hospital workers. We randomly divided respondents into derivation and validation samples. We applied exploratory factor analysis to responses in the derivation sample. We used those results to create scales in the validation sample, which we subjected to multitrait analysis (MTA). We identified nine constructs, three organizational factors, two unit factors, three individual factors, and one additional factor. Constructs demonstrated substantial convergent and discriminant validity in the MTA. Cronbach's alpha coefficients ranged from 0.50 to 0.89. It is possible to measure key salient features of hospital safety climate using a valid and reliable 38-item survey and appropriate hospital sample sizes. This instrument may be used in further studies to better understand the impact of safety climate on patient safety outcomes.

Preliminary design of smart fuel

International Nuclear Information System (INIS)

Kim, Y.; Ha, D.; Park, S.; Nahm, K.; Lee, K.; Kim, J.

2007-01-01

SMART (System-integrated Modular Advanced Reactor) is a novel light water rector with a modular, integral primary system configuration. This concept has been developing a 660 MWt by Korean Nuclear Power Industry Group with KAERI. SMART is being developed for use as an energy source for small-scale power generation and seawater desalination. Although the design of SMART is based on the current pressurized water reactor technology, new technologies such as enhanced safety, and passive safety have been applied, and system simplification and modularization, innovations in manufacturing and installation technologies have been implemented culminating in a design that has enhanced safety and economy, and is environment -friendly. In this paper described the preliminary design of the nuclear Fuel for this SMART, the design concept and the characteristics of SMART Fuel. In specially this paper describe the optimization of grid span adjustment to improve the thermal performance of the SMART Fuel as well as to improve the seismic resistance performance of the SMART Fuel, it is not easy to improve the both performance simultaneously because of design parameter of each performance inversely proportional. SMART Fuel enable to extra-long extended fuel cycle length and resistance of proliferation, enhanced safety, improved economics and reduced nuclear waste

Study of fieldbus technology confiability when applied in a Sterilization plant control and safety systems

International Nuclear Information System (INIS)

Karma, D.; Sampa, M.H.O.; Rela, P.R.

2001-01-01

Several sterilization processes have been used in these years for treatment of countless products. Some processes use high temperatures, thermal shocks and chemical agents. With the discovery of the ionizing radiation and its posterior technological developments turned possible the application of that process, in 1960, also in the sterilization, denominated radiation sterilization. This process became also applied in another areas of health and industrial as food conservation, gemstones enhancement and others. The radiation sterilization requests an effective control and it needs a high level of safety. The commercial use of the computers applied in industrial automation provides and the domain of new technologies in this field provides news applications then new designs now is possible. The Fieldbus technology, a new digital communication protocol, like a Local Area Network, can be an alternative in the cobalt-60 irradiation plant. This paper show preliminary study about confiability in systems using Fieldbus technology. This technology was simulated in sterilization plant control and safety systems and the fail probability was quantified using Fail Tree Analysis Method. Fieldbus technology can be used in sterilization plants because the confiability in this systems is like PLCs and relays systems, was the conclusion

Center for Maritime Safety and Health Studies

Data.gov (United States)

Federal Laboratory Consortium — Established in November 2015, the Center for Maritime Safety and Health Studies (CMSHS) promotes safety and health for all maritime workers, including those employed...

Flood risk and economically optimal safety targets for coastal flood defense systems

NARCIS (Netherlands)

Dupuits, E.J.C.; Schweckendiek, T.

2015-01-01

A front defense can improve the reliability of a rear defense in a coastal flood defense system. The influence of this interdependency on the accompanying economically optimal safety targets of both front and rear defense is investigated. The results preliminary suggest that the optimal safety level

A Sensitivity Study for an Evaluation of Input Parameters Effect on a Preliminary Probabilistic Tsunami Hazard Analysis

Energy Technology Data Exchange (ETDEWEB)

Rhee, Hyun-Me; Kim, Min Kyu; Choi, In-Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Sheen, Dong-Hoon [Chonnam National University, Gwangju (Korea, Republic of)

2014-10-15

The tsunami hazard analysis has been based on the seismic hazard analysis. The seismic hazard analysis has been performed by using the deterministic method and the probabilistic method. To consider the uncertainties in hazard analysis, the probabilistic method has been regarded as attractive approach. The various parameters and their weight are considered by using the logic tree approach in the probabilistic method. The uncertainties of parameters should be suggested by analyzing the sensitivity because the various parameters are used in the hazard analysis. To apply the probabilistic tsunami hazard analysis, the preliminary study for the Ulchin NPP site had been performed. The information on the fault sources which was published by the Atomic Energy Society of Japan (AESJ) had been used in the preliminary study. The tsunami propagation was simulated by using the TSUNAMI{sub 1}.0 which was developed by Japan Nuclear Energy Safety Organization (JNES). The wave parameters have been estimated from the result of tsunami simulation. In this study, the sensitivity analysis for the fault sources which were selected in the previous studies has been performed. To analyze the effect of the parameters, the sensitivity analysis for the E3 fault source which was published by AESJ was performed. The effect of the recurrence interval, the potential maximum magnitude, and the beta were suggested by the sensitivity analysis results. Level of annual exceedance probability has been affected by the recurrence interval.. Wave heights have been influenced by the potential maximum magnitude and the beta. In the future, the sensitivity analysis for the all fault sources in the western part of Japan which were published AESJ would be performed.

Study on Fusion Safety Infrastructure using ISAM

International Nuclear Information System (INIS)

Oh, Kyemin; Kang, Myungsuk; Heo, Gyunyoung; Kim, Hyoungchan

2013-01-01

The regulation of nuclear facilities have checked and managed safety throughout the entire process from design, construction, operation and decommissioning. Also, the same meaning as the regulatory requirements and design requirements, it will be important indicators for detailed design of K-DEMO. K-DEMO has many uncertainties because it is in conceptual design phase. Also, there is no reference material because demonstration scale fusion power plants were not operated yet in overseas. So, hazard that threaten the integrity of K-DEMO have to be defined preferentially to define regulatory or design requirements. This study proposed method that educe regulatory or design requirements and introduce web-based cloud infrastructure to perform renewal and sharing of information related with safety that is required in the study rapidly as a part of the R and D program funded by National Fusion Research Institute of Korea (NFRI). We have been performing QSR and PIRT in accordance with development of fusion DEMO plant, and preparing OPT, PSA and DPA for regulation requirements. This study introduces our recent research activities about ISAM for fusion and CCI built for expert and extant safety related information. Unlike fission, nuclear fusion's safety goal is non-evacuation of the public during an accident. To satisfy this goal not only various safety issues should be analyzed, but safety objectives, regulatory requirements, and design variables should also be established in detailed design phase. The web-based cloud infrastructure proposed in this paper will be able to offer input data of future studies and, it is expected to contribute on general and technical safety principles for national fusion power plant technology plan

Preliminary Study of a Piston Pump for Cryogenic Fluids

Science.gov (United States)

Biermann, Arnold E.; Kohl, Robert C.

1959-01-01

Preliminary data are presented covering the performance of a low-speed, five-cylinder piston pump designed for handling boiling hydrogen. This pump was designed for a flow of 55 gallons per minute at 240 rpm with a discharge pressure of 135 pounds per square inch. Tests were made using JP-4 fuel, liquid nitrogen, and liquid hydrogen. Pump delivery and endurance characteristics were satisfactory for the range of operation covered. In connection with the foregoing pump development, the cavitation characteristics of a preliminary visual model, glass-cylinder pump and of a simple reciprocating disk were studied. Subcooling of approximately 0.60 F was obtained from the cavitation produced by reciprocating a disk in boiling nitrogen and in boiling water. The subcooling obtained in a similar manner with liquid hydrogen was somewhat less.

The first symposium of Research Center for Radiation Safety, NIRS. Perspective of future studies of radiation safety

International Nuclear Information System (INIS)

Shimo, Michikuni

2002-03-01

This paper summarizes presentations given in the title symposium, held at the Conference Room of National Institute of Radiological Sciences (NIRS) on November 29 and 30, 2001. Contained are Introductory remarks: Basic presentations concerning exposure dose in man; Environmental levels of radiation and radioactivity, environmental radon level and exposure dose, and radiation levels in the specific environment (like in the aircraft): Special lecture (biological effects given by space environment) concerning various needs for studies of radiation safety; Requirement for open investigations, from the view of utilization, research and development of atomic energy, from the clinical aspect, and from the epidemiological aspect: Special lecture (safety in utilization of atomic energy and radiation-Activities of Nuclear Safety Commission of Japan) concerning present state and perspective of studies of radiation safety; Safety of radiation and studies of biological effects of radiation-perspective, and radiation protection and radiation safety studies: Studies in the Research Center for Radiation Safety; Summary of studies in the center, studies of the biological effects of neutron beam, carcinogenesis by radiation and living environmental factors-complicated effects, and studies of hereditary effects: Panel discussion (future direction of studies of radiation safety for the purpose of the center's direction): and concluding remarks. (N.I.)

Safety overview of the National Ignition Facility

International Nuclear Information System (INIS)

Brereton, S.J.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.

1996-01-01

The National Ignition Facility (NIF) is a proposed US Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium- tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low

Preliminary study on the establishment of the radionuclide declaration methods for radionuclides in LILW radioactive waste

International Nuclear Information System (INIS)

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

2003-01-01

The preliminary study on declaration methods has been done for each radionuclide in LILW radwaste drum in Korean NPPs. View from the preliminary establishment of radio nuclide declaration methods, The selection of assessment target nuclide through the qualitative method and preliminary criteria for routine declaration methods in each radio nuclide was derived. First of all, selection criteria and preliminary assessment method for each target radionuclide was surveyed and investigated. And, the selection criteria and selected the target radio nuclides from the basis on criteria was derived. And the preliminary suggestion about the declaration methods for each target radio nuclide was established

Updated safety analysis of ITER

Energy Technology Data Exchange (ETDEWEB)

Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2011-10-15

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Updated safety analysis of ITER

International Nuclear Information System (INIS)

Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

2011-01-01

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Online gaming dependency: a preliminary study in China.

Science.gov (United States)

Peng, Wei; Liu, Ming

2010-06-01

Based on theories and previous studies on problematic Internet use, we propose a model to better understand the contributors to and consequences of online gaming dependency. A preliminary study was conducted through a survey of online gamers in China. The results of path analysis found that maladaptive cognitions, shyness, and depression are positively related to online gaming dependency. Online gaming dependency was also positively related to different types of negative life outcomes. The findings of this study have implications for the prevention and treatment of addictive online gaming.

Preliminary thermal and stress analysis of the SINQ window

International Nuclear Information System (INIS)

Heidenreich, G.

1991-01-01

Preliminary results of a finite element analysis for the SINQ proton beam window are presented. Temperatures and stresses are calculated in an axisymmetric model. As a result of these calculations, the H 2 O-cooled window (safety window) could be redesigned in such a way that plastic deformation resulting from excessive stress in some areas is avoided. (author)

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)

Preliminary study on the dye removal efficacy of immobilized marine ...

African Journals Online (AJOL)

Preliminary study on the dye removal efficacy of immobilized marine and freshwater microalgal beads from textile wastewater. SD Kumar, P Santhanam, R Nandakumar, S Anath, B Balaji Prasath, A Shenbaga Devi, S Jeyanthi, T Jayalakshima, P Ananthi ...

Pre-feasibility study for final disposal of radioactive waste. Disposal concepts. Main report

International Nuclear Information System (INIS)

Andersen, L.; Skov, C.; Kueter, A.; Schepper, L.; Gottberg Roemer, H.; Refsgaard, A.; Utko, M.; Kristiansen, Torben

2011-05-01

This prefeasibility study is part of the overall process related to the decision on placement and design of a repository for the Danish low and medium level radioactive waste primarily from the facilities at Risoe. The prefeasibility study encompasses the preliminary design of a number of repository types based on the overall types set out in the 'Parliamentary decision' together with a preliminary safety assessment of these repository types based on their possible placement in a set of typical Danish geologies. The report consists of three parts. Part I is the descriptive part containing information on the waste to be disposed of, the potential conditioning (packaging) possibilities for the waste before placement in a repository, the suggested preliminary design of the different repository types, and the suggested visual appearance of the repository. Part II is the assessment part. It contains an introduction to the concepts used in the preliminary safety assessment, which encompasses: the assessment of potential long term impact and the assessment of possible accidental incidents. The division of the preliminary safety assessment in to these two categories has several reasons. One is that the criteria to which impact is to be compared are different for the two types of impact, another is that while the possible variation in the long term impact is primarily due to the possible variation in the parameters influencing the impact, the impact from accidental incidents is governed by the probability of the occurrence of these incidents and the potential consequence of the impact, which calls for a different assessment approach. Since the suggestions for packaging of the different waste types is a result of both types of assessments, part II also contains a description of these suggestions based on the preliminary safety assessments. Finally part II contains the costs related to the different types of repositories and the suggested packaging. Part III of the report

Pre-feasibility study for final disposal of radioactive waste. Disposal concepts. Main report

Energy Technology Data Exchange (ETDEWEB)

Andersen, L.; Skov, C.; Kueter, A.; Schepper, L.; Gottberg Roemer, H.; Refsgaard, A.; Utko, M.; Kristiansen, Torben (COWI A/S, Kgs. Lyngby (Denmark))

2011-05-15

This prefeasibility study is part of the overall process related to the decision on placement and design of a repository for the Danish low and medium level radioactive waste primarily from the facilities at Risoe. The prefeasibility study encompasses the preliminary design of a number of repository types based on the overall types set out in the 'Parliamentary decision' together with a preliminary safety assessment of these repository types based on their possible placement in a set of typical Danish geologies. The report consists of three parts. Part I is the descriptive part containing information on the waste to be disposed of, the potential conditioning (packaging) possibilities for the waste before placement in a repository, the suggested preliminary design of the different repository types, and the suggested visual appearance of the repository. Part II is the assessment part. It contains an introduction to the concepts used in the preliminary safety assessment, which encompasses: the assessment of potential long term impact and the assessment of possible accidental incidents. The division of the preliminary safety assessment in to these two categories has several reasons. One is that the criteria to which impact is to be compared are different for the two types of impact, another is that while the possible variation in the long term impact is primarily due to the possible variation in the parameters influencing the impact, the impact from accidental incidents is governed by the probability of the occurrence of these incidents and the potential consequence of the impact, which calls for a different assessment approach. Since the suggestions for packaging of the different waste types is a result of both types of assessments, part II also contains a description of these suggestions based on the preliminary safety assessments. Finally part II contains the costs related to the different types of repositories and the suggested packaging. Part III of the

Lessons learned from measuring safety culture: an Australian case study.

Science.gov (United States)

Allen, Suellen; Chiarella, Mary; Homer, Caroline S E

2010-10-01

adverse events in maternity care are relatively common but often avoidable. International patient safety strategies advocate measuring safety culture as a strategy to improve patient safety. Evidence suggests it is necessary to fully understand the safety culture of an organisation to make improvements to patient safety. this paper reports a case study examining the safety culture in one maternity service in Australia and considers the benefits of using surveys and interviews to understand safety culture as an approach to identify possible strategies to improve patient safety in this setting. the study took place in one maternity service in two public hospitals in NSW, Australia. Concurrently, both hospitals were undergoing an organisational restructure which was part of a major health reform agenda. The priorities of the reform included improving the quality of care and patient safety; and, creating a more efficient health system by reducing administration inefficiencies and duplication. a descriptive case study using three approaches: the safety culture was identified to warrant improvement across all six safety culture domains. There was reduced infrastructure and capacity to support incident management activities required to improve safety, which was influenced by instability from the organisational restructure. There was a perceived lack of leadership at all levels to drive safety and quality and improving the safety culture was neither a key priority nor was it valued by the organisation. the safety culture was complex as was undertaking this study. We were unable to achieve a desired 60% response rate highlighting the limitations of using safety culture surveys in isolation as a strategy to improve safety culture. Qualitative interviews provided greater insight into the factors influencing the safety culture. The findings of this study provide evidence of the benefits of including qualitative methods with quantitative surveys when examining safety culture

Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

Energy Technology Data Exchange (ETDEWEB)

1994-09-01

This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.

Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

International Nuclear Information System (INIS)

1994-09-01

This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included

Radiology workstation for mammography: preliminary observations, eyetracker studies, and design

Science.gov (United States)

Beard, David V.; Johnston, Richard E.; Pisano, Etta D.; Hemminger, Bradley M.; Pizer, Stephen M.

1991-07-01

For the last four years, the UNC FilmPlane project has focused on constructing a radiology workstation facilitating CT interpretations equivalent to those with film and viewbox. Interpretation of multiple CT studies was originally chosen because handling such large numbers of images was considered to be one of the most difficult tasks that could be performed with a workstation. The authors extend the FilmPlane design to address mammography. The high resolution and contrast demands coupled with the number of images often cross- compared make mammography a difficult challenge for the workstation designer. This paper presents the results of preliminary work with workstation interpretation of mammography. Background material is presented to justify why the authors believe electronic mammographic workstations could improve health care delivery. The results of several observation sessions and a preliminary eyetracker study of multiple-study mammography interpretations are described. Finally, tentative conclusions of what a mammographic workstation might look like and how it would meet clinical demand to be effective are presented.

Preliminary study of soil permeability properties using principal component analysis

Science.gov (United States)

Yulianti, M.; Sudriani, Y.; Rustini, H. A.

2018-02-01

Soil permeability measurement is undoubtedly important in carrying out soil-water research such as rainfall-runoff modelling, irrigation water distribution systems, etc. It is also known that acquiring reliable soil permeability data is rather laborious, time-consuming, and costly. Therefore, it is desirable to develop the prediction model. Several studies of empirical equations for predicting permeability have been undertaken by many researchers. These studies derived the models from areas which soil characteristics are different from Indonesian soil, which suggest a possibility that these permeability models are site-specific. The purpose of this study is to identify which soil parameters correspond strongly to soil permeability and propose a preliminary model for permeability prediction. Principal component analysis (PCA) was applied to 16 parameters analysed from 37 sites consist of 91 samples obtained from Batanghari Watershed. Findings indicated five variables that have strong correlation with soil permeability, and we recommend a preliminary permeability model, which is potential for further development.

PTSD and Impaired Eye Expression Recognition: A Preliminary Study

Science.gov (United States)

Schmidt, Jakob Zeuthen; Zachariae, Robert

2009-01-01

This preliminary study examined whether posttraumatic stress disorder (PTSD) was related to difficulties in identifying the mental states of others in a group of refugees. Sixteen Bosnian refugees, referred to treatment in an outpatient treatment center for survivors of torture and war-related trauma in Denmark (CETT), were compared to 16 non-PTSD…

Preliminary safety analysis for key design features of KALIMER with breakeven core

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

2001-06-01

KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

Safety and Effectiveness of Highly Active Antiretroviral Therapy in Treatment-Naïve HIV Patients: Preliminary Findings of a Cohort Event Monitoring Study in Belarus.

Science.gov (United States)

Setkina, Svetlana; Dotsenko, Marina; Bondar, Sviatlana; Charnysh, Iryna; Kuchko, Alla; Kaznacheeva, Alena; Kozorez, Elena; Dodaleva, Alena; Rossa, Natalia

2015-04-01

Antiretroviral drugs have well-documented evidence-based favorable benefit-risk ratios. Although various studies have investigated and characterized the safety profile of antiretroviral medicines, there are a limited number of studies evaluating the safety of first-line antiretroviral therapy (ART) in patients with a specific co-morbidity. A cohort event monitoring (CEM) study of the safety and effectiveness of antiretroviral medicines in a target population that has a significant level of co-morbidities (chronic infectious diseases, peripheral blood cytopenias) was implemented. The aim was to evaluate the safety profile of the highly active ART (HAART) in the target population and subpopulations with risk factors, to optimize the monitoring and decision-making procedure for subgroups of patients with specific types of co-morbidity, and to implement a more vigilant approach to therapy management in risk groups of patients. Prospective observational CEM was implemented among HAART-naïve HIV-positive patients at four clinical sites from December 2012. Eligible patients were those starting first-line HAART. Close medical supervision of all enrolled patients, with regular clinical and laboratory monitoring, was provided by healthcare professionals within 1 year after commencement of therapy. Standardized forms were used for data collection on initial and subsequent visits. All objective or subjective deviations in condition (events) were assessed for a causal relationship with ART, and for severity, seriousness, reversibility, preventability, and pre-existing risk factors in the case of adverse drug reactions (ADRs). A total of 518 HAART-naïve HIV-positive patients were enrolled in the CEM study. Of these patients, 65% (337) experienced one or several ADRs related to one or more components of HAART. Most of the ADRs reported were non-serious, expected, common (very common), transient (correctable), or reversible. The most common were hematotoxic, hepatotoxic, and

Spirituality and the Events of September 11: A Preliminary Study

Science.gov (United States)

Briggs, Michele Kielty; Apple, Kevin J.; Aydlett, Ann E.

2004-01-01

Personal crises have been associated with spiritual growth. Sparked by the global response to the crisis of September 11, 2001, this study examined the relationship of spirituality and the September 11 tragedy using a sample of convenience from a college student population. This preliminary study used an experimental design to examine various…

Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2016-05-15

To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

Novel modular natural circulation BWR design and safety evaluation

International Nuclear Information System (INIS)

Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

2015-01-01

Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

Safety and cost evaluation of nuclear waste management

International Nuclear Information System (INIS)

Vieno, T.; Hautojaervi, A.; Korhonen, R.

1989-11-01

The report introduces the results of the nuclear waste management safety and cost evaluation research carried out in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1984-1988. The emphasis is on the description of the state-of-art of performance and cost evaluation methods. The report describes VTT's most important assessment models. Development, verification and validation of the models has largely taken place within international projects, including the Stripa, HYDROCOIN, INTRACOIN, INTRAVAL, PSACOIN and BIOMOVS projects. Furthermore, VTT's other laboratories are participating in the Natural Analogue Working Group,k the CHEMVAL project and the CoCo group. Resent safety analyses carried out in the Nuclear Engineering Laboratory include a concept feasibility study of spent fuel disposal, safety analyses for the Preliminary Safety Analysis Reports (PSAR's) of the repositories to be constructed for low and medium level operational reactor waste at the Olkiluoto and Loviisa power plants as well as safety analyses of disposal of decommissioning wastes. Appendix 1 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail

Targeting paretic propulsion to improve poststroke walking function: a preliminary study.

Science.gov (United States)

Awad, Louis N; Reisman, Darcy S; Kesar, Trisha M; Binder-Macleod, Stuart A

2014-05-01

To determine the feasibility and safety of implementing a 12-week locomotor intervention targeting paretic propulsion deficits during walking through the joining of 2 independent interventions, walking at maximal speed on a treadmill and functional electrical stimulation of the paretic ankle musculature (FastFES); to determine the effects of FastFES training on individual subjects; and to determine the influence of baseline impairment severity on treatment outcomes. Single group pre-post preliminary study investigating a novel locomotor intervention. Research laboratory. Individuals (N=13) with locomotor deficits after stroke. FastFES training was provided for 12 weeks at a frequency of 3 sessions per week and 30 minutes per session. Measures of gait mechanics, functional balance, short- and long-distance walking function, and self-perceived participation were collected at baseline, posttraining, and 3-month follow-up evaluations. Changes after treatment were assessed using pairwise comparisons and compared with known minimal clinically important differences or minimal detectable changes. Correlation analyses were run to determine the correlation between baseline clinical and biomechanical performance versus improvements in walking speed. Twelve of the 13 subjects that were recruited completed the training. Improvements in paretic propulsion were accompanied by improvements in functional balance, walking function, and self-perceived participation (each Pstudy of this promising locomotor intervention for persons poststroke. Copyright © 2014 American Congress of Rehabilitation Medicine. Published by Elsevier Inc. All rights reserved.

The Preliminary Review for the Cross-Cutting Issues in the US Nuclear Regulatory Inspection Framework

International Nuclear Information System (INIS)

Lee, Yong Suk; Jung, Dae Wook; Cho, Nam Chul

2008-01-01

The research for the development of risk-informed and performance-based regulatory inspection is ongoing in KINS. In the USNRC, the cross-cutting issue is one of the main components the risk-informed and performance-based regulatory inspection process as shown in figure 1, which is named as ROP (Reactor Oversight Process). The following three cross-cutting areas implicitly affect all of the safety cornerstones in ROP. In this study, the preliminary review for the inspection practices of cross-cutting issues in the US and Korean safety regulatory system were performed. The elements of the cross-cutting issues were recently modified to emphasize the importance of safety culture, and the graded approach was applied for the inspection of cross-cutting issues in USNRC. The graded approach for the inspection of cross-cutting issues will be also needed to Korean safety regulatory system in the future

The Preliminary Review for the Cross-Cutting Issues in the US Nuclear Regulatory Inspection Framework

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Suk; Jung, Dae Wook [Future and Challenges Inc., Seoul (Korea, Republic of); Cho, Nam Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2008-10-15

The research for the development of risk-informed and performance-based regulatory inspection is ongoing in KINS. In the USNRC, the cross-cutting issue is one of the main components the risk-informed and performance-based regulatory inspection process as shown in figure 1, which is named as ROP (Reactor Oversight Process). The following three cross-cutting areas implicitly affect all of the safety cornerstones in ROP. In this study, the preliminary review for the inspection practices of cross-cutting issues in the US and Korean safety regulatory system were performed. The elements of the cross-cutting issues were recently modified to emphasize the importance of safety culture, and the graded approach was applied for the inspection of cross-cutting issues in USNRC. The graded approach for the inspection of cross-cutting issues will be also needed to Korean safety regulatory system in the future.

Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

International Nuclear Information System (INIS)

Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

2013-01-01

SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

Do provisions to advance chemical facility safety also advance chemical facility security? An analysis of possible synergies

DEFF Research Database (Denmark)

Hedlund, Frank Huess

2012-01-01

The European Commission has launched a study on the applicability of existing chemical industry safety provisions to enhancing security of chemical facilities covering the situation in 18 EU Member States. This paper reports some preliminary analytical findings regarding the extent to which exist...

Preliminary safety information document for the standard MHTGR. Volume 4

Energy Technology Data Exchange (ETDEWEB)

None

1986-01-01

This report contains information concerning: operational radionuclide control; occupational radiation protection, conduct of operations; initial test program; safety analysis; technical specifications; and quality assurance. (JDB)

Pharmacokinetics and Preliminary Safety of Pod-Intravaginal Rings Delivering the Monoclonal Antibody VRC01-N for HIV Prophylaxis in a Macaque Model.

Science.gov (United States)

Zhao, Chunxia; Gunawardana, Manjula; Villinger, Francois; Baum, Marc M; Remedios-Chan, Mariana; Moench, Thomas R; Zeitlin, Larry; Whaley, Kevin J; Bohorov, Ognian; Smith, Thomas J; Anderson, Deborah J; Moss, John A

2017-07-01

The broadly neutralizing antibody (bNAb) VRC01, capable of neutralizing 91% of known human immunodeficiency virus type 1 (HIV-1) isolates in vitro , is a promising candidate microbicide for preventing sexual HIV infection when administered topically to the vagina; however, accessibility to antibody-based prophylactic treatment by target populations in sub-Saharan Africa and other underdeveloped regions may be limited by the high cost of conventionally produced antibodies and the limited capacity to manufacture such antibodies. Intravaginal rings of the pod design (pod-IVRs) delivering Nicotiana -manufactured VRC01 (VRC01-N) over a range of release rates have been developed. The pharmacokinetics and preliminary safety of VRC01-N pod-IVRs were evaluated in a rhesus macaque model. The devices sustained VRC01-N release for up to 21 days at controlled rates, with mean steady-state VRC01-N levels in vaginal fluids in the range of 10 2 to 10 3 μg g -1 being correlated with in vitro release rates. No adverse safety indications were observed. These findings indicate that pod-IVRs are promising devices for the delivery of the candidate topical microbicide VRC01-N against HIV-1 infection and merit further preclinical evaluation. Copyright © 2017 American Society for Microbiology.

Do provisions to advance chemical facility safety also advance chemical facility security? - An analysis of possible synergies

OpenAIRE

Hedlund, Frank Huess

2012-01-01

The European Commission has launched a study on the applicability of existing chemical industry safety provisions to enhancing security of chemical facilities covering the situation in 18 EU Member States. This paper reports some preliminary analytical findings regarding the extent to which existing provisions that have been put into existence to advance safety objectives due to synergy effects could be expected advance security objectives as well.The paper provides a conceptual definition of...

Angra-1 probabilistic safety study-phase B

International Nuclear Information System (INIS)

Fernandes Filho, T.L.; Gibelli, S.M.O.

1988-05-01

This study represents the Phase B of the Angra-1 Probabilistic Safety Study and is the the final report prepared for the IAEA under Research Contract No. 3423/R2/RB. The three main items covered in this report are the establishment of interim safety goals, analysis of Angra-1 operational experience and development of emergency procedures to address severe accidents. For establishment of interim safety goals a methodology for calculating consequences and risks associated to the Angra-1 operation was developed based on the available data and codes. The proposed safety goals refer to the individual risk of early fatality for people living in the vicinity of the plant, colective risk of cancer fatalities for people living near the plant, the propobability of core melt occurrence and the probability of dominant accident sequences. (author) [pt

Behavioral integrity for safety, priority of safety, psychological safety, and patient safety : a team-level study

NARCIS (Netherlands)

Leroy, H.; Dierynck, B.; Anseel, F.; Simons, T.; Halbesleben, J.R.; McCaughey, D.; Savage, G.T.; Sels, L.

2012-01-01

This article clarifies how leader behavioral integrity for safety helps solve follower's double bind between adhering to safety protocols and speaking up about mistakes against protocols. Path modeling of survey data in 54 nursing teams showed that head nurse behavioral integrity for safety

A preliminary study on the relevancy of sustainable building design ...

African Journals Online (AJOL)

This preliminary study aims to explore the relationship between sustainable building design paradigms and commercial property depreciation, to assist in the understanding of sustainable building design impact towards commercial building value and rental de employs the qualitative method and analyses valuers' current ...

A Preliminary Study toward Consistent Soil Moisture from AMSR2

NARCIS (Netherlands)

Parinussa, R.M.; Holmes, T.R.H.; Wanders, N.; Dorigo, W.A.; de Jeu, R.A.M.

2015-01-01

A preliminary study toward consistent soil moisture products from the Advanced Microwave Scanning Radiometer 2 (AMSR2) is presented. Its predecessor, the Advanced Microwave Scanning Radiometer for Earth Observing System (AMSR-E), has providedEarth scientists with a consistent and continuous global

French regulatory approach to establishing the safety case for ageing NPP's

International Nuclear Information System (INIS)

Delage, M.

1994-06-01

The French regulatory procedures make provision for three main stages in the safety assessment of nuclear power plants. The first stage ends up with the construction licence and focuses on the assessment of the preliminary safety report. The second stage makes it possible to issue the fuel loading approval following evaluation of the provisional safety report. The third stage permits to declare the start of normal operation of the installation. The procedure, the tests and the assessment forming the overall strategy for safety regulations are described in detail. (R.P.)

A study on safety climate at nuclear power plants

International Nuclear Information System (INIS)

Fukui, Hirokazu; Yoshida, Michio; Yoshiyama, Naohiro

2001-01-01

In the current study, we define safety climate as an organizational environment that induces members of the organization to give consideration to safety or take safety actions. It is of utmost importance that people holding managerial positions in an organization have a good understanding of the characteristics of the safety climate of the organization and implement safety promotion activities effectively. In the current research, we studied the rating scales and the characteristics of a safety climate. A survey was conducted, targeting technical engineers who belong to the three power stations of Kansai Electric Power Co., Inc. The questionnaire mainly consisted of questions concerning safety measures taken by individuals and questions concerning safety measures taken by the organization, to which the individuals belong. As a result of a factor analysis of the responses, we extracted five factors, namely, 'confidence in knowledge and skill', attitude of supervisors,' 'safety education in workplace', 'clarity of tasks' and 'safety confirmation/report'. In studying the rating scales of the safety climate, we selected five items from each of the above five factors, and used the total scores of the ratings of the five items as scores of each factor. Then, we examined the correlation between scores of personal factors and scores of organizational environment factors. We treated the scores of safety confirmation/report' and 'confidence in knowledge and skill', which are personal factors, as criterion variables, and the scores of 'attitude of supervisors', 'safety education in workplace' and 'clarity of tasks', which are organizational environment factors, as predictor variables. As a result, we found that levels of 'safety confirmation/report' and 'confidence in knowledge and skill' can be deduced from the scores of 'attitude of supervisors', 'safety education in workplace' and 'clarity of tasks.' Hence, we have decided to use these three organizational environment

Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

Preliminary study and Identification of insects' species of forensic ...

African Journals Online (AJOL)

The proper identification of the insect and arthropod species of forensic importance is the most crucial element in the field of forensic entomology. The main objective in this study was the identification of insects' species of forensic importance in Urmia (37°, 33 N. and 45°, 4, 45 E.) and establishment of a preliminary ...

Preliminary study on chicken feather protein-based wood adhesives

Science.gov (United States)

Zehui Jiang; Daochun Qin; Chung-Yun Hse; Monlin Kuo; Zhaohui Luo; Ge Wang; Yan Yu

2008-01-01

The objective of this preliminary study was to partially replace phenol in the synthesis of phenol-formaldehyde resin with feather protein. Feather protein–based resins, which contained one part feather protein and two parts phenol, were formulated under the conditions of two feather protein hydrolysis methods (with and without presence of phenol during...

Criticality safety basics, a study guide

Energy Technology Data Exchange (ETDEWEB)

V. L. Putman

1999-09-01

This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates.

Criticality safety basics, a study guide

International Nuclear Information System (INIS)

Putman, V.L.

1999-01-01

This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates

A preliminary conceptual design study for Korean fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

2013-10-15

Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

Preliminary Study on Non-Fatal Occupational Injury among Operational Workers in Malaysia Palm Oil Mill

Directory of Open Access Journals (Sweden)

Ruslan Rumaizah

2017-01-01

Full Text Available Non-fatal occupational injury had becoming major global concern and its consequences to safety and health would be heavily burdening. The aim of this preliminary study was to investigate the distribution of non-fatal occupational injury among specific group of workers in palm oil mill and to acknowledge potential factors of injury causation. A questionnaire survey was designed to assess injury involvement during the employment period among operational workers of palm oil mill located in Southern Peninsular Malaysia. Thirty three (n= 33 workers volunteered and completed the questionnaire. Prevalence of injury among palm oil mill workers was 39.4% with sprain and burn were the common types of injury reported. Press Plant workers reported to have high cases of injuries. Majority of workers (78.8% stated noise was the main hazard in the palm oil mill, followed by heat hazard. Press Plant was identified as one of the specific risk group in palm oil mill based on the rate of injury occurrences reported by the workers. Exposure to noise hazard was estimated to be one of the potential factors of injury causation and further study should be made to measure the risk of hazard.

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)

Safety certification of airborne software: An empirical study

International Nuclear Information System (INIS)

Dodd, Ian; Habli, Ibrahim

2012-01-01

Many safety-critical aircraft functions are software-enabled. Airborne software must be audited and approved by the aerospace certification authorities prior to deployment. The auditing process is time-consuming, and its outcome is unpredictable, due to the criticality and complex nature of airborne software. To ensure that the engineering of airborne software is systematically regulated and is auditable, certification authorities mandate compliance with safety standards that detail industrial best practice. This paper reviews existing practices in software safety certification. It also explores how software safety audits are performed in the civil aerospace domain. The paper then proposes a statistical method for supporting software safety audits by collecting and analysing data about the software throughout its lifecycle. This method is then empirically evaluated through an industrial case study based on data collected from 9 aerospace projects covering 58 software releases. The results of this case study show that our proposed method can help the certification authorities and the software and safety engineers to gain confidence in the certification readiness of airborne software and predict the likely outcome of the audits. The results also highlight some confidentiality issues concerning the management and retention of sensitive data generated from safety-critical projects.

A novel topical association with zinc oxide, chamomile and aloe vera extracts - stability and safety studies

Directory of Open Access Journals (Sweden)

Catarina Reis

2015-12-01

Full Text Available Currently, natural products show an enormous potential for pharmaceutical and cosmetic industries. The goals of this study were to formulate and to characterise a novel combination of natural products. Formulations were 1 an oil-in-water emulsion, 2 a water-in-oil emulsion and 3 a cleansing solution. Zinc oxide was chosen as an active ingredient due to its healing properties, and chamomile and aloe vera extracts were chosen due to their antiseptic, anti-inflammatory and tissue regenerating properties. Organoleptic characteristics, pH, viscosity and in vitro efficacy for the most common bacteria and yeast of human skin were evaluated. Preliminary and accelerated stability studies and safety tests were also performed. All optimized products were stable, smooth in texture, effective against bacteria and yeast, and safe, justifying further studies. Results suggest that these novel products might be a promising source of natural compounds with soothing and regenerative properties for skin care.

An evaluation of the uranium mine radiation safety course

International Nuclear Information System (INIS)

1984-07-01

The report evaluates the Uranium Mine Radiation Safety Course focussing on the following areas: effectivenss of the course; course content; instructional quality; course administration. It notes strengths and weaknesses in these areas and offers preliminary recommendations for future action

Preliminary safety examination on thorium molten-salt nuclear energy synergetics

International Nuclear Information System (INIS)

Furukawa, Kazuo

1991-01-01

The new global fission industry for the next century should keep a strong public acceptance, which means to ensure an enough rational safety feature not only in the engineering issue but also in all issues of integral fuel-cycle system. In this sense, the safety characteristics of the Thorium Molten-Salt Nuclear Energy Synergetic System (THORIMS-NES) is widely examined relating with the several aspects of environmental (including resources, radio-waste, etc.) social (including anti-nuclear proliferation and terrorism, etc), basic technological, engineering, institutional, and economical aspects. From this examination it seems that this system is verified as one of the most promising measures of North-South problem, greenhouse effect, etc in the world. (author). 11 refs., 3 figs., 5 tabs

Quality management of pharmacology and safety pharmacology studies

DEFF Research Database (Denmark)

Spindler, Per; Seiler, Jürg P

2002-01-01

to safety pharmacology studies, and, when indicated, to secondary pharmacodynamic studies, does not influence the scientific standards of studies. However, applying formal GLP standards will ensure the quality, reliability and integrity of studies, which reflect sound study management. It is important...... to encourage a positive attitude among researchers and academics towards these lines, whenever possible. GLP principles applied to the management of non-clinical safety studies are appropriate quality standards when studies are used in the context of protecting public health, and these quality standards...... of pharmacology studies (ICH S7A): primary pharmacodynamic, secondary pharmacodynamic and safety pharmacology studies, and guidance on the quality standards (expectations for GLP conformity) for these study types have been provided. Primary pharmacodynamic studies are the only study types that are fully exempt...

Technique for unit testing of safety software verification and validation

International Nuclear Information System (INIS)

Li Duo; Zhang Liangju; Feng Junting

2008-01-01

The key issue arising from digitalization of the reactor protection system for nuclear power plant is how to carry out verification and validation (V and V), to demonstrate and confirm the software that performs reactor safety functions is safe and reliable. One of the most important processes for software V and V is unit testing, which verifies and validates the software coding based on concept design for consistency, correctness and completeness during software development. The paper shows a preliminary study on the technique for unit testing of safety software V and V, focusing on such aspects as how to confirm test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed here was successfully used in the work of unit testing on safety software of a digital reactor protection system. (authors)

A study on safety climate at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fukui, Hirokazu [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Yoshida, Michio; Yoshiyama, Naohiro [Japan Institute for Group Dynamics, Fukuoka (Japan)

2001-09-01

In the current study, we define safety climate as an organizational environment that induces members of the organization to give consideration to safety or take safety actions. It is of utmost importance that people holding managerial positions in an organization have a good understanding of the characteristics of the safety climate of the organization and implement safety promotion activities effectively. In the current research, we studied the rating scales and the characteristics of a safety climate. A survey was conducted, targeting technical engineers who belong to the three power stations of Kansai Electric Power Co., Inc. The questionnaire mainly consisted of questions concerning safety measures taken by individuals and questions concerning safety measures taken by the organization, to which the individuals belong. As a result of a factor analysis of the responses, we extracted five factors, namely, 'confidence in knowledge and skill', attitude of supervisors,' 'safety education in workplace', 'clarity of tasks' and 'safety confirmation/report'. In studying the rating scales of the safety climate, we selected five items from each of the above five factors, and used the total scores of the ratings of the five items as scores of each factor. Then, we examined the correlation between scores of personal factors and scores of organizational environment factors. We treated the scores of safety confirmation/report' and 'confidence in knowledge and skill', which are personal factors, as criterion variables, and the scores of 'attitude of supervisors', 'safety education in workplace' and 'clarity of tasks', which are organizational environment factors, as predictor variables. As a result, we found that levels of 'safety confirmation/report' and 'confidence in knowledge and skill' can be deduced from the scores of 'attitude of supervisors', 'safety

Design and safety studies on an EFIT core with CERMET fuel

International Nuclear Information System (INIS)

Chen, Xue-Nong; Rineiski, Andrei; Liu, Ping; Maschek, Werner; Matzerath Boccaccini, Claudia; Gabrielli, Fabrizio; Sobolev, Vitaly

2008-01-01

Within the EUROTRANS Programme a European Facility for Industrial Transmutation (EFIT) is under development. This paper deals with the design and safety analyses of an EFIT core with Mo-matrix based CERMET fuel. A three zone core design was developed, which satisfies the EFIT general and specific requirements. The fuel/matrix ratio in each zone is determined for a suitable subcritical level at a k eff of about 0.97 and a total form factor around 1.5. The Pu/MA ratio also determines the transmutation rate and the burn-up characteristics, ranging between 46/54 at% to 40/60 at% for optimizing the reactivity swing and the MA transmutation efficiency. Based on the preliminary core design, safety calculations are performed with SIMMER-III for three types of transient: the unprotected loss of flow (ULOF), the unprotected transient of over power (UTOP) and the unprotected blockage accident (UBA). It can be shown that in the CERMET core the fuel and clad design limits are not violated under the conditions of ULOF and UTOP. In the UBA case, pin failures will happen and lead to a local voiding and reactivity insertion, but a fuel sweep-out process leads to a power reduction and restricts the core degradation. (authors)

LM-OSL from single grains of quartz: A preliminary study

DEFF Research Database (Denmark)

Bulur, E.; Duller, G.A.T.; Solongo, S.

2002-01-01

the easy-to-bleach component, those with only the hard-to-bleach component, and those exhibiting all components. The results of this preliminary study show that LM-OSL experiments carried out at the single grain level may give important insights into the luminescence properties observed when viewing...

Management of construction safety at KKNPP site

International Nuclear Information System (INIS)

Khare, P.K.

2016-01-01

Construction is considered as one of the most hazardous activities owing to the number of accidents and injuries. At KKNPP, management of industrial safety has been envisaged since the preliminary stage of construction planning, including design aspects. The governing principles of safety management are evolved from the Factories Act, 1948, the Atomic Energy(Factories) Rules, 1996, AERB safety guidelines on Control of works (2011) and Corporate HSE policy of NPCIL (2014). Numerous risk assessment and hazard control measures are adopted consistently to ensure a safe work environment during the construction, which includes Job Hazard Analysis, work permit through Computerized Maintenance Management System, safety procedures, exclusive safety training facility for the contractor's workmen, safety motivational measures, safety surveillance and reporting through Safety Related Deficiencies Management System. Assessment of efficacy of safety management system is continuously done through safety audits and observations are being circulated and discussed in committee meetings. Fire safety is also being taken care of since inception of project work. Well-equipped fire station with trained fire fighters was made available since the beginning as per AERB safety standard on fire protection system for Nuclear facilities. Fire prevention measures specific to the work are implemented during all activities. (author)

Children: Oklahoma's Investment in Tomorrow '96. Preliminary Report: Agency Budget by Cabinet.

Science.gov (United States)

Oklahoma Commission on Children and Youth, Oklahoma City.

This report presents preliminary Oklahoma state agency budget summaries for all programs serving children in the Departments of Administration, Agriculture, Commerce, Education, Energy, Health and Human Services, Human Resources, Safety and Security, Tourism and Recreation, and Veterans Affairs. The budget figures are organized by cabinet and…

Safety Risk Knowledge Elicitation in Support of Aeronautical R and D Portfolio Management: A Case Study

Science.gov (United States)

Shih, Ann T.; Ancel, Ersin; Jones, Sharon Monica; Reveley, Mary S.; Luxhoj, James T.

2012-01-01

Aviation is a problem domain characterized by a high level of system complexity and uncertainty. Safety risk analysis in such a domain is especially challenging given the multitude of operations and diverse stakeholders. The Federal Aviation Administration (FAA) projects that by 2025 air traffic will increase by more than 50 percent with 1.1 billion passengers a year and more than 85,000 flights every 24 hours contributing to further delays and congestion in the sky (Circelli, 2011). This increased system complexity necessitates the application of structured safety risk analysis methods to understand and eliminate where possible, reduce, and/or mitigate risk factors. The use of expert judgments for probabilistic safety analysis in such a complex domain is necessary especially when evaluating the projected impact of future technologies, capabilities, and procedures for which current operational data may be scarce. Management of an R&D product portfolio in such a dynamic domain needs a systematic process to elicit these expert judgments, process modeling results, perform sensitivity analyses, and efficiently communicate the modeling results to decision makers. In this paper a case study focusing on the application of an R&D portfolio of aeronautical products intended to mitigate aircraft Loss of Control (LOC) accidents is presented. In particular, the knowledge elicitation process with three subject matter experts who contributed to the safety risk model is emphasized. The application and refinement of a verbal-numerical scale for conditional probability elicitation in a Bayesian Belief Network (BBN) is discussed. The preliminary findings from this initial step of a three-part elicitation are important to project management practitioners as they illustrate the vital contribution of systematic knowledge elicitation in complex domains.

Study on 'Safety qualification of process computers used in safety systems of nuclear power plants'

International Nuclear Information System (INIS)

Bertsche, K.; Hoermann, E.

1991-01-01

The study aims at developing safety standards for hardware and software of computer systems which are increasingly used also for important safety systems in nuclear power plants. The survey of the present state-of-the-art of safety requirements and specifications for safety-relevant systems and, additionally, for process computer systems has been compiled from national and foreign rules. In the Federal Republic of Germany the KTA safety guides and the BMI/BMU safety criteria have to be observed. For the design of future computer-aided systems in nuclear power plants it will be necessary to apply the guidelines in [DIN-880] and [DKE-714] together with [DIN-192]. With the aid of a risk graph the various functions of a system, or of a subsystem, can be evaluated with regard to their significance for safety engineering. (orig./HP) [de

NPP Temelin safety analysis reports and PSA status

International Nuclear Information System (INIS)

Mlady, O.

1999-01-01

To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

Efficacy and safety of sublingual tablets of house dust mite allergen extracts in adults with allergic rhinitis

NARCIS (Netherlands)

Bergmann, Karl-Christian; Demoly, Pascal; Worm, Margitta; Fokkens, Wytske J.; Carrillo, Teresa; Tabar, Ana I.; Nguyen, Hélène; Montagut, Armelle; Zeldin, Robert K.

2014-01-01

Preliminary studies have suggested the efficacy of sublingual tablets of house dust mite (HDM) extracts in adults with allergic rhinitis. We sought to assess the efficacy and safety of 2 doses of HDM sublingual tablets over 1 treatment year and the subsequent immunotherapy-free year. Adults with

Safety and efficacy of aneurysm treatment with WEB

DEFF Research Database (Denmark)

Pierot, Laurent; Costalat, Vincent; Moret, Jacques

2016-01-01

OBJECT WEB is an innovative intrasaccular treatment for intracranial aneurysms. Preliminary series have shown good safety and efficacy. The WEB Clinical Assessment of Intrasaccular Aneurysm Therapy (WEBCAST) trial is a prospective European trial evaluating the safety and efficacy of WEB in wide......-neck bifurcation aneurysms. METHODS Patients with wide-neck bifurcation aneurysms for which WEB treatment was indicated were included in this multicentergood clinical practices study. Clinical data including adverse events and clinical status at 1 and 6 months were collected and independently analyzed by a medical....... RESULTS Ten European neurointerventional centers enrolled 51 patients with 51 aneurysms. Treatment with WEB was achieved in 48 of 51 aneurysms (94.1%). Adjunctive implants (coils/stents) were used in 4 of 48 aneurysms (8.3%). Thromboembolic events were observed in 9 of 51 patients (17.6%), resulting...

The PEC reactor. Safety analysis: Detailed reports

Energy Technology Data Exchange (ETDEWEB)

1988-01-01

In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

Pancreatic head cryosurgery: safety and efficiency in vivo--a pilot study.

Science.gov (United States)

Li, Jialiang; Zhou, Liang; Chen, Jibing; Wu, Binghui; Zeng, Jianying; Fang, Gang; Deng, Chunjuan; Huang, Shengquan; Yao, Fei; Chen, Zhixian; Leng, Yin; Deng, Min; Deng, Chunmei; Zhang, Bo; Zhou, Gang; He, Lihua; Liao, Maoxin; Chiu, David; Niu, Lizhi; Zuo, Jiansheng; Xu, Kecheng

2012-11-01

Pancreatic cancer is the fourth leading cause of cancer-related death. Cryosurgery has emerged as a promising new technique for treatment. Although 80% of pancreatic cancers are located in the pancreatic head, no research has been conducted on the safety and efficacy of cryosurgery for these tumors. Two groups of Tibetan miniature pigs (n = 4 per group) underwent cryosurgery to the pancreatic head with either the deep freezing protocol (100% argon output) or shallow freezing protocol (10% argon output), and compared to sham-operated pigs. Serum inflammatory factors and amylase increased during the 5 days after cryoablation in both groups but acute pancreatitis did not occur. Adhesions were observed between the pancreatic head and adjacent organs, and only minor trauma was caused to the stomach, duodenum, small intestine, and liver. Ice balls with a radius of 0.5 cm beyond the tumor edge were sufficient to cause complete necrosis of the pancreatic tissue, and decreased the degree of cold injury to surrounding tissues. Shallow freezing protocol seemed to be safer than, and just as effective as, the deep freezing protocol. This preliminary study suggests that cryosurgery could potentially be an effective treatment of cancer of the pancreatic head.

Environmental Regulation and Food Safety: Studies of Protection ...

International Development Research Centre (IDRC) Digital Library (Canada)

Environmental Regulation and Food Safety: Studies of Protection and Protectionism. Book cover Environmental Regulation and Food Safety: Studies of Protection and Protectionism. Directeur(s) : Veena Jha. Maison(s) d'édition : Edward Elgar, IDRC. 1 janvier 2006. ISBN : 184542512X. 250 pages. e-ISBN : 155250185X.

A Preliminary Study for Safety Shutter design to Protect Streaming of Residual Radiation Passing through Beamline in Pre-Separator Room of ISOL

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Woo; Kim, Do Hyun; Kim, Song Hyun; Shin, Chang Ho; Nam, Shin Woo [KAERI, Daejeon (Korea, Republic of)

2016-05-15

RAON is a heavy ion accelerator under construction by the Institute for Basic Science (IBS) in Korea. As one part of the RAON accelerator, ISOL is a facility to generate and separate rare isotopes for various experiments. In ISOL facility, isotopes generated from the reaction between 70 MeV proton beam and UC{sub 2} target are transferred to pre-separator room. Almost all isotopes accumulated in slit of pre-separator except specific isobars, which are set for experiments. Residual radiations are generated from accumulated isotopes because these isotopes are unstable. Streaming of residual radiation by the beamline is weak point for radiation shielding design. In this study, safety shutter was designed. Residual radiation generated from accumulated isotopes at slit of pre-separator was estimated using following conditions: (1) the isotopes generated by proton-target reactions are accumulated at slit with 10 % accumulation rate; (2) it was assumed that the radioactive isotopes are uniformly distributed in the cylindrical slit which have 1 cm height and 15 diameter. To design optimized safety shutter, following steps were performed: (1) thickness and diameter of the bulk shield material were evaluated to optimize safety shutter material; (2) additional shielding structure was proposed using dose contribution of each additional shielding wall.

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

Energy Technology Data Exchange (ETDEWEB)

Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2013-06-15

Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

Nitrogen-system safety study: Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

1982-07-01

The Department of Energy has primary responsibility for the safety of operations at DOE-owned nuclear facilities. The guidelines for the analysis of credible accidents are outlined in DOE Order 5481.1. DOE has requested that existing plant facilities and operations be reviewed for potential safety problems not covered by standard industrial safety procedures. This review is being conducted by investigating individual facilities and documenting the results in Safety Study Reports which will be compiled to form the Existing Plant Final Safety Analysis Report which is scheduled for completion in September, 1984. This Safety Study documents the review of the Plant Nitrogen System facilities and operations and consists of Section 4.0, Facility and Process Description, and Section 5.0, Accident Analysis, of the Final Safety Analysis Report format. The existing nitrogen system consists of a Superior Air Products Company Type D Nitrogen Plant, nitrogen storage facilities, vaporization facilities and a distribution system. The system is designed to generate and distribute nitrogen gas used in the cascade for seal feed, buffer systems, and for servicing equipment when exceptionally low dew points are required. Gaseous nitrogen is also distributed to various process auxiliary buildings. The average usage is approximately 130,000 standard cubic feet per day

Preliminary study of radium-contaminated soils

Energy Technology Data Exchange (ETDEWEB)

Healy, J.W.; Rodgers, J.C.

1978-10-01

A preliminary study was made of the potential radiation exposures to people from radium-226 contamination in the soil in order to provide guidance on limits to be applied in decontaminating land. Pathways included were inhalation of radium from resuspension; ingestion of radium with foods; external gamma radiation from radium daughters; inhalation of radon and daughter, both in the open air and in houses; and the intake of /sup 210/Pb and /sup 210/Po from both inhalation and ingestion. The depth of the contaminated layer is of importance for external exposure and especially for radon emanation. The most limiting pathway was found to be emanation of the radon into buildings with limiting values comparable to those found naturally in many areas.

Preliminary study of radium-contaminated soils

International Nuclear Information System (INIS)

Healy, J.W.; Rodgers, J.C.

1978-10-01

A preliminary study was made of the potential radiation exposures to people from radium-226 contamination in the soil in order to provide guidance on limits to be applied in decontaminating land. Pathways included were inhalation of radium from resuspension; ingestion of radium with foods; external gamma radiation from radium daughters; inhalation of radon and daughter, both in the open air and in houses; and the intake of 210 Pb and 210 Po from both inhalation and ingestion. The depth of the contaminated layer is of importance for external exposure and especially for radon emanation. The most limiting pathway was found to be emanation of the radon into buildings with limiting values comparable to those found naturally in many areas

FEP catalogue for the VSG. Documentation. Report on the work package 7. Preliminary safety analysis Gorleben (VSG); FEP-Katalog fuer die VSG. Dokumentation. Bericht zum Arbeitspaket 7. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

Energy Technology Data Exchange (ETDEWEB)

Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Behlau, Joachim [Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover (Germany); Beuth, Thomas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)] [and others

2012-06-15

The report is a compendium of the FEP (features, events, processes) data base that was developed in the frame of the preliminary safety analysis Gorleben (VSG). For each FEP issue the information includes the following subchapters: definition, general information and examples, status at the site, site-specific impacts, temporal restriction, conditional incidence rate, effects on subsidiary systems, adverse effects on the function of the initial barriers, justification, direct dependencies, open questions, references.

French regulatory approach to establishing the safety case for ageing NPP`s

Energy Technology Data Exchange (ETDEWEB)

Delage, M.

1994-06-15

The French regulatory procedures make provision for three main stages in the safety assessment of nuclear power plants. The first stage ends up with the construction licence and focuses on the assessment of the preliminary safety report. The second stage makes it possible to issue the fuel loading approval following evaluation of the provisional safety report. The third stage permits to declare the start of normal operation of the installation. The procedure, the tests and the assessment forming the overall strategy for safety regulations are described in detail. (R.P.).

The X-Ray Pebble Recirculation Experiment (X-PREX): Facility Description, Preliminary Discrete Element Method Simulation Validation Studies, and Future Test Program

International Nuclear Information System (INIS)

Laufer, Michael R.; Bickel, Jeffrey E.; Buster, Grant C.; Krumwiede, David L.; Peterson, Per F.

2014-01-01

This paper presents a facility description, preliminary results, and future test program of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Preliminary experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. Finally, this paper discusses additional studies in progress relevant to the design and analysis of pebble bed reactor cores including pebble recirculation in cylindrical core geometries and evaluation of forces on shut down blades inserted directly into a packed pebble bed. (author)

On Safety Management. A Frame of Reference for Studies of Safety Management with Examples From Non-Nuclear Contexts of Relevance for Nuclear Safety

International Nuclear Information System (INIS)

Svensson, Ola; Salo, Ilkka; Allwin, Pernilla

2004-11-01

A good knowledge about safety management from risk technologies outside the area of nuclear power may contribute to both broaden the perspectives on safety management in general, and point at new opportunities for improving safety measures within the nuclear industry. First, a theoretical framework for the study of safety management in general is presented, followed by three case studies on safety management from different non-nuclear areas with potential relevance for nuclear safety. The chapters are written as separate reports and can be read independently of each other. The nuclear industry has a long experience about the management of risky activities, involving all the stages from planing to implementation, both on a more generalized level and in the specific branches of activities (management, administration, operation, maintenance, etc.). Here, safety management is a key concept related to these areas of activities. Outside the field of nuclear power there exist a number of different non-nuclear risk technologies, each one with their own specific needs and experiences about safety management. The differences between the areas consist partly of the different experiences caused by the different technologies. Besides using own experiences in safety practices within the own areas of activities, it may be profitable to take advantage in knowledge and experiences from one area and put it in practice in another area. In order to facilitate knowledge transfer from one technological area to another it may be possible to adapt a common theoretical model, for descriptions and explanations, to the different technologies. Such a model should admit that common denominators for safety management across the areas might be identified and described with common concepts. Systems theory gives the opportunity to not only create models that are descriptive for events within the limits of a given technology, but also to generate knowledge that can be transferred to other

The relationship between team climate and interprofessional collaboration: Preliminary results of a mixed methods study.

Science.gov (United States)

Agreli, Heloise F; Peduzzi, Marina; Bailey, Christopher

2017-03-01

Relational and organisational factors are key elements of interprofessional collaboration (IPC) and team climate. Few studies have explored the relationship between IPC and team climate. This article presents a study that aimed to explore IPC in primary healthcare teams and understand how the assessment of team climate may provide insights into IPC. A mixed methods study design was adopted. In Stage 1 of the study, team climate was assessed using the Team Climate Inventory with 159 professionals in 18 interprofessional teams based in São Paulo, Brazil. In Stage 2, data were collected through in-depth interviews with a sample of team members who participated in the first stage of the study. Results from Stage 1 provided an overview of factors relevant to teamwork, which in turn informed our exploration of the relationship between team climate and IPC. Preliminary findings from Stage 2 indicated that teams with a more positive team climate (in particular, greater participative safety) also reported more effective communication and mutual support. In conclusion, team climate provided insights into IPC, especially regarding aspects of communication and interaction in teams. Further research will provide a better understanding of differences and areas of overlap between team climate and IPC. It will potentially contribute for an innovative theoretical approach to explore interprofessional work in primary care settings.

A preliminary study on growth response of broiler finishers fed ...

African Journals Online (AJOL)

A preliminary study on growth response of broiler finishers fed processed mottle Mucuna beans ( Mucuna pruriens var. utilis ) ... They were fed diets (20% CP, 13 MJME/kg) incorporating 0%, 5% and 10% processed mottle “Mucuna” beans. A completely randomized design was used. Feed and water were supplied and ...

Studying the Safety Impact of Autonomous Vehicles Using Simulation-Based Surrogate Safety Measures

Directory of Open Access Journals (Sweden)

Mark Mario Morando

2018-01-01

Full Text Available Autonomous vehicle (AV technology has advanced rapidly in recent years with some automated features already available in vehicles on the market. AVs are expected to reduce traffic crashes as the majority of crashes are related to driver errors, fatigue, alcohol, or drugs. However, very little research has been conducted to estimate the safety impact of AVs. This paper aims to investigate the safety impacts of AVs using a simulation-based surrogate safety measure approach. To this end, safety impacts are explored through the number of conflicts extracted from the VISSIM traffic microsimulator using the Surrogate Safety Assessment Model (SSAM. Behaviours of human-driven vehicles (HVs and AVs (level 4 automation are modelled within the VISSIM’s car-following model. The safety investigation is conducted for two case studies, that is, a signalised intersection and a roundabout, under various AV penetration rates. Results suggest that AVs improve safety significantly with high penetration rates, even when they travel with shorter headways to improve road capacity and reduce delay. For the signalised intersection, AVs reduce the number of conflicts by 20% to 65% with the AV penetration rates of between 50% and 100% (statistically significant at p<0.05. For the roundabout, the number of conflicts is reduced by 29% to 64% with the 100% AV penetration rate (statistically significant at p<0.05.

Contribution of operating feedback to probabilistic safety studies

International Nuclear Information System (INIS)

Guio, J.M. de; Lannoy, A.

1992-03-01

This paper presents the method used for PWR unit operation feedback analysis and its contribution to probabilistic safety studies. The targets were as follows: - use of failure data banks to assess reliability parameters, - use of event data banks to identify and quantify main system initiating events, - determination of a standard operating profile. These studies, performed in the context of nuclear power plant safety programs, prove useful not only to safety engineers but also to equipment experts, designers, operators and maintenance specialists. They constitute basic data for studies in all these areas or the departure point for new investigations. (authors). 3 figs., 3 tabs., 3 refs

A Disposable Tear Glucose Biosensor-Part 4: Preliminary Animal Model Study Assessing Efficacy, Safety, and Feasibility.

Science.gov (United States)

La Belle, Jeffrey T; Engelschall, Erica; Lan, Kenneth; Shah, Pankti; Saez, Neil; Maxwell, Stephanie; Adamson, Teagan; Abou-Eid, Michelle; McAferty, Kenyon; Patel, Dharmendra R; Cook, Curtiss B

2014-01-01

A prototype tear glucose (TG) sensor was tested in New Zealand white rabbits to assess eye irritation, blood glucose (BG) and TG lag time, and correlation with BG. A total of 4 animals were used. Eye irritation was monitored by Lissamine green dye and analyzed using image analysis software. Lag time was correlated with an oral glucose load while recording TG and BG readings. Correlation between TG and BG were plotted against one another to form a correlation diagram, using a Yellow Springs Instrument (YSI) and self-monitoring of blood glucose as the reference measurements. Finally, TG levels were calculated using analytically derived expressions. From repeated testing carried over the course of 12 months, little to no eye irritation was detected. TG fluctuations over time visually appeared to trace the same pattern as BG with an average lag times of 13 minutes. TG levels calculated from the device current measurements ranged from 4 to 20 mg/dL and correlated linearly with BG levels of 75-160 mg/dL (TG = 0.1723 BG = 7.9448 mg/dL; R 2 = .7544). The first steps were taken toward preliminary development of a sensor for self-monitoring of tear glucose (SMTG). No conjunctival irritation in any of the animals was noted. Lag time between TG and BG was found to be noticeable, but a quantitative modeling to correlate lag time in this study is unnecessary. Measured currents from the sensors and the calculated TG showed promising correlation to BG levels. Previous analytical bench marking showed BG and TG levels consistent with other literature. © 2014 Diabetes Technology Society.

Criticality safety studies at VTT Energy

International Nuclear Information System (INIS)

Roine, T.; Anttila, M.

1995-01-01

At VTT Energy a compact reactor physics calculation system is applied in many kind of problems. Generation of group constants for static and dynamic core calculations, flux and dose rate calculations as well as criticality safety studies are performed basically with the same codes. In the presentation a short overview of the wide variety of criticality safety problems analyzed at VTT Energy is given. The calculation system with some illustrative examples is also described. (12 refs., 1 tab.)

In Vitro Studies and Preliminary Mathematical Model for Jet Fuel and Noise Induced Auditory Impairment

Science.gov (United States)

2015-06-01

of JP-8 and a Fischer- Tropsch synthetic jet fuel following subacute inhalation exposure in rats. Toxicol Sci 116(1): 239-248. Gallinat, J...AFRL-RH-WP-TR-2015-0084 IN VITRO STUDIES AND PRELIMINARY MATHEMATICAL MODEL FOR JET FUEL AND NOISE INDUCED AUDITORY IMPAIRMENT...April 2014 – September 2014 4. TITLE AND SUBTITLE In Vitro Studies and Preliminary Mathematical Model for Jet Fuel and Noise Induced Auditory

The German nuclear power plant safety study

International Nuclear Information System (INIS)

1979-01-01

With this study a new approach has been chosen, taking nuclear power plants as an example to assess and to describe the risks arising from the use of modern technology, including those hazards emanating from the rather hypothetical possibility of occurrence of very serious accidents. Following the definition of basic concepts and methods to be applied in risk assessment studied, as well as a brief account of the design and operating mode of nuclear power plants with PWRs', accidents and failures to be considered in a safety study are described. Using the course-of-event and fault tree analysis, the probability of fission product release as a consequence of failures in safety systems or of core meltdown is evaluated. Subsequently, the theoretical model for assessment of reactor accident consequences is presented, discussing such aspects as the dispersion of radioactivity in the atmosphere, the radiation dose model, safety and countermeasures, the model for the evaluation of health hazards as well as methods and calculations for estimating the reliability of risk assessments together with the remaining uncertainties. In an appendix to this study, the analyses presented in the study are discussed in the light of the TMI-2 event. This safety study showing the possibilities of detecting, keeping in check and minimizing harmful effects, can be regarded as a contribution to a better understanding of our modern, highly industrialised society, and eventually to an improvement of the quality of life. (GL) 891 GL/GL 892 MB [de

Buffer erosion: An overview of concepts and potential safety consequences

International Nuclear Information System (INIS)

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

2010-11-01

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

Buffer erosion: An overview of concepts and potential safety consequences

Energy Technology Data Exchange (ETDEWEB)

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

2010-11-15

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

On Safety Management. A Frame of Reference for Studies of Safety Management with Examples From Non-Nuclear Contexts of Relevance for Nuclear Safety

Energy Technology Data Exchange (ETDEWEB)

Svensson, Ola; Salo, Ilkka; Allwin, Pernilla (Risk Analysis, Social and Decision Research Unit, Dept. of Psychology, Stockholm Univ., Stockholm (Sweden))

2004-11-15

A good knowledge about safety management from risk technologies outside the area of nuclear power may contribute to both broaden the perspectives on safety management in general, and point at new opportunities for improving safety measures within the nuclear industry. First, a theoretical framework for the study of safety management in general is presented, followed by three case studies on safety management from different non-nuclear areas with potential relevance for nuclear safety. The chapters are written as separate reports and can be read independently of each other. The nuclear industry has a long experience about the management of risky activities, involving all the stages from planing to implementation, both on a more generalized level and in the specific branches of activities (management, administration, operation, maintenance, etc.). Here, safety management is a key concept related to these areas of activities. Outside the field of nuclear power there exist a number of different non-nuclear risk technologies, each one with their own specific needs and experiences about safety management. The differences between the areas consist partly of the different experiences caused by the different technologies. Besides using own experiences in safety practices within the own areas of activities, it may be profitable to take advantage in knowledge and experiences from one area and put it in practice in another area. In order to facilitate knowledge transfer from one technological area to another it may be possible to adapt a common theoretical model, for descriptions and explanations, to the different technologies. Such a model should admit that common denominators for safety management across the areas might be identified and described with common concepts. Systems theory gives the opportunity to not only create models that are descriptive for events within the limits of a given technology, but also to generate knowledge that can be transferred to other

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Study of extraterrestrial disposal of radioactive wastes. Part 3: Preliminary feasibility screening study of space disposal of the actinide radioactive wastes with 1 percent and 0.1 percent fission product contamination

Science.gov (United States)

Hyland, R. E.; Wohl, M. L.; Finnegan, P. M.

1973-01-01

A preliminary study was conducted of the feasibility of space disposal of the actinide class of radioactive waste material. This waste was assumed to contain 1 and 0.1 percent residual fission products, since it may not be feasible to completely separate the actinides. The actinides are a small fraction of the total waste but they remain radioactive much longer than the other wastes and must be isolated from human encounter for tens of thousands of years. Results indicate that space disposal is promising but more study is required, particularly in the area of safety. The minimum cost of space transportation would increase the consumer electric utility bill by the order of 1 percent for earth escape and 3 percent for solar escape. The waste package in this phase of the study was designed for normal operating conditions only; the design of next phase of the study will include provisions for accident safety. The number of shuttle launches per year required to dispose of all U.S. generated actinide waste with 0.1 percent residual fission products varies between 3 and 15 in 1985 and between 25 and 110 by 2000. The lower values assume earth escape (solar orbit) and the higher values are for escape from the solar system.

Survey of electronic safety systems in accelerator applications

International Nuclear Information System (INIS)

Mahoney, K.

1997-01-01

This paper presents the preliminary results and analysis of a comprehensive survey of the implementation of accelerator safety interlock systems from over 30 international labs. At the present time there is not a self consistent means to evaluate both the experiences and level of protection provided by electronic safety interlock systems. This research is intended to analyze the strength and weaknesses of several different types of interlock system implementation methodologies. Research, medical, and industrial accelerators are compared. Thomas Jefferson National Accelerator Facility (TJNAF) was one of the first large particle accelerators to implement a safety interlock system using programmable logic controllers. Since that time all of the major new U.S. accelerator construction projects plan to use some form of programmable electronics as part of a safety interlock system in some capacity

User-Centered Collaborative Design and Development of an Inpatient Safety Dashboard.

Science.gov (United States)

Mlaver, Eli; Schnipper, Jeffrey L; Boxer, Robert B; Breuer, Dominic J; Gershanik, Esteban F; Dykes, Patricia C; Massaro, Anthony F; Benneyan, James; Bates, David W; Lehmann, Lisa S

2017-12-01

Patient safety remains a key concern in hospital care. This article summarizes the iterative participatory development, features, functions, and preliminary evaluation of a patient safety dashboard for interdisciplinary rounding teams on inpatient medical services. This electronic health record (EHR)-embedded dashboard collects real-time data covering 13 safety domains through web services and applies logic to generate stratified alerts with an interactive check-box function. The technological infrastructure is adaptable to other EHR environments. Surveyed users perceived the tool as highly usable and useful. Integration of the dashboard into clinical care is intended to promote communication about patient safety and facilitate identification and management of safety concerns. Copyright © 2017 The Joint Commission. All rights reserved.

Review of studies on criticality safety evaluation and criticality experiment methods

International Nuclear Information System (INIS)

Naito, Yoshitaka; Yamamoto, Toshihiro; Misawa, Tsuyoshi; Yamane, Yuichi

2013-01-01

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS. (author)

Preliminary safety assessment of C-8 xylitol monoester and xylitol phosphate esters.

Science.gov (United States)

Silveira, J E P S; Pereda, M C V; Nogueira, C; Dieamant, G; Cesar, C K M; Assanome, K M; Silva, M S; Torello, C O; Queiroz, M L S; Eberlin, S

2016-02-01

Most of the cosmetic compounds with preservative properties available in the market pose some risks concerning safety, such as the possibility of causing sensitization. Due to the fact that there are few options, the proper development of new molecules with this purpose is needed. Xylitol is a natural sugar, and the antimicrobial properties of xylitol-derived compounds have already been described in the literature. C-8 xylitol monoester and xylitol phosphate esters may be useful for the development of skincare products. As an initial screen for safety of chemicals, the combination of in silico methods and in vitro testing can aid in prioritizing resources in toxicological investigations while reducing the ethical and monetary costs that are related to animal and human testing. This study was designed to evaluate the safety of C-8 xylitol monoester and xylitol phosphate esters regarding carcinogenicity, mutagenicity, skin and eye irritation/corrosion and sensitization through alternative methods. For the initial safety assessment, quantitative structure-activity relationship methodology was used. The prediction of the parameters carcinogenicity/mutagenicity, skin and eye irritation/corrosion and sensitization was generated from the chemical structure. The analysis also comprised physical-chemical properties, Cramer rules, threshold of toxicological concern and Michael reaction. In silico results of candidate molecules were compared to 19 compounds with preservative properties that are available in the market. Additionally, in vitro tests (Ames test for mutagenicity, cytotoxicity and phototoxicity tests and hen's egg test--chorioallantoic membrane for irritation) were performed to complement the evaluation. In silico evaluation of both molecules presented no structural alerts related to eye and skin irritation, corrosion and sensitization, but some alerts for micronucleus and carcinogenicity were detected. However, by comparison, C-8 xylitol monoester, xylitol

Dataset for Phase I randomized clinical trial for safety and tolerability of GET 73 in single and repeated ascending doses including preliminary pharmacokinetic parameters.

Science.gov (United States)

Haass-Koffler, Carolina L; Goodyear, Kimberly; Long, Victoria M; Tran, Harrison H; Loche, Antonella; Cacciaglia, Roberto; Swift, Robert M; Leggio, Lorenzo

2017-12-01

The data in this article outline the methods used for the administration of GET 73 in the first time-in-human manuscript entitled "Phase I randomized clinical trial for the safety, tolerability and preliminary pharmacokinetics of the mGluR5 negative allosteric modulator GET 73 following single and repeated doses in healthy male volunteers" (Haass-Koffler et al., 2017) [1]. Data sets are provided in two different manners. The first series of tables provided includes procedural information about the experiments conducted. The next series of tables provided includes Pharmacokinetic (PK) parameters for GET 73 and its main metabolite MET 2. This set of data is comprised by two experiments: Experiment 1 references a single ascending dose administration of GET 73 and Experiment 2 references a repeated ascending dose administration of GET 73.

On safety management and nuclear safety - A frame of reference for studies of safety management with examples from non-nuclear contects of relevance for nuclear safety

International Nuclear Information System (INIS)

Svenson, O.; Allwin, P.; Salo, I.

2004-03-01

The report includes three case studies of safety management. The studies are presented as chapters, but are written in a format that makes them easy to read separately. Two of the studies cover regulators (the Swedish Civil Aviation Safety Authority, Luftfartsinspektionen) and the Norwegian Petroleum Directorate) and one a regulated activity/industry (a car manufacturer, Volvo Car). The introduction outlines a living system framework and relates this to concepts used in organizational management. The report concludes with some findings with potential relevance for safety management in the nuclear power domain. In the next phase of the work, the regulated counterparts of the regulators here will be investigated in addition to a fourth case study of a regulated activity/industry. (au)

A study of passive safety conditions for fast reactor core

International Nuclear Information System (INIS)

Shimizu, Akinao

1991-01-01

A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

Study of industry safety management

International Nuclear Information System (INIS)

Park, Pil Su

1987-06-01

This book deals with general remarks, industrial accidents, statistics of industrial accidents, unsafe actions, making machinery and facilities safe, safe activities, having working environment safe, survey of industrial accidents and analysis of causes, system of safety management and operations, safety management planning, safety education, human engineering such as human-machines system, system safety, and costs of disaster losses. It lastly adds individual protective equipment and working clothes including protect equipment for eyes, face, hands, arms and feet.

Towards a Formal Basis for Modular Safety Cases

Science.gov (United States)

Denney, Ewen; Pai, Ganesh

2015-01-01

Safety assurance using argument-based safety cases is an accepted best-practice in many safety-critical sectors. Goal Structuring Notation (GSN), which is widely used for presenting safety arguments graphically, provides a notion of modular arguments to support the goal of incremental certification. Despite the efforts at standardization, GSN remains an informal notation whereas the GSN standard contains appreciable ambiguity especially concerning modular extensions. This, in turn, presents challenges when developing tools and methods to intelligently manipulate modular GSN arguments. This paper develops the elements of a theory of modular safety cases, leveraging our previous work on formalizing GSN arguments. Using example argument structures we highlight some ambiguities arising through the existing guidance, present the intuition underlying the theory, clarify syntax, and address modular arguments, contracts, well-formedness and well-scopedness of modules. Based on this theory, we have a preliminary implementation of modular arguments in our toolset, AdvoCATE.

A comparison of safety belt use between commercial and noncommercial light-vehicle occupants.

Science.gov (United States)

Eby, David W; Fordyce, Tiffani A; Vivoda, Jonathon M

2002-05-01

The purpose of this study was to conduct an observational survey of safety belt use to determine the use rate of commercial versus noncommercial light-vehicle occupants. Observations were conducted on front-outboard vehicle occupants in eligible commercial and noncommercial vehicles in Michigan (i.e.. passenger cars, vans/minivans, sport-utility vehicles, and pickup trucks). Commercial vehicles that did not fit into one of the four vehicle type categories, such as tractor-trailers, buses, or heavy trucks, were not included in the survey. The study found that the restraint use rate for commercial light-vehicle occupants was 55.8% statewide. The statewide safety belt use rate for commercial light-vehicles was significantly lower than the rate of 71.2% for noncommercial light-vehicles. The safety belt use rate for commercial vehicles was also significantly different as a function of region, vehicle type, seating position, age group, and road type. The results provide important preliminary data about safety belt use in commercial versus noncommercial light-vehicles and indicate that further effort is needed to promote safety belt use in the commercial light-vehicle occupant population. The study also suggests that additional research is required in order to develop effective programs that address low safety belt use in the commercial light-vehicle occupant population.

Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

International Nuclear Information System (INIS)

Setiadipura, T.; Zuhair; Irwanto, D.

2015-01-01

The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

Directory of Open Access Journals (Sweden)

T. Setiadipura

2015-04-01

Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

Corporate Social Disclosures in Southeast Asia: A Preliminary Study

Directory of Open Access Journals (Sweden)

Juniati Gunawan

2012-12-01

Full Text Available The issue of Corporate Social Disclosure (CSD has been growing remarkably both in business and academic world.  Inevitably, this topic is also exposed in Southeast Asia, a big region that plays important role in global economic issue. Applying a content analysis method, this paper aims to provide preliminary findings in CSD practices throughout the companies‟ annual reports in 2007 and 2008 for countries located in Southeast Asia.  Samples were selected for listed and unlisted various type of industries, based on the information availability internet searching. The sample collection and the subjectivity during the content analysis process are the limitations in conducting this study. In general, the results show that „human resources‟ are the main information disclosed, while in contrast, „energy‟ is the main least issue disclosed in the annual reports.  However, the findings need to be interpreted with considerations since there are limited in samples. Basically, the outcomes support the major prior studies and enhancing the discussion of CSD conducting in developing countries, while at the same time describing some countries which obtained very limited in exposures. To respond the vast increasing issues of CSD practice, this preliminary study has provided a basis to see the role of every country in CSR reporting and how they could support the sustainability development globally.

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-08-15

This report documents a study with the objective of enhancing safety culture in the Swedish nuclear power industry. A primary objective of this study was to ensure that the latest thinking on human factors principles was being recognised and applied by nuclear power operators as a means of ensuring optimal safety performance. The initial phase of the project was conducted as a pilot study, involving the senior management group at one Swedish nuclear power-producing site. The pilot study enabled the project methodology to be validated after which it was repeated at other Swedish nuclear power industry sites, providing a broad-ranging analysis of opportunities across the industry to enhance safety culture. The introduction to this report contains an overview of safety culture, explains the background to the project and sets out the project rationale and objectives. The methodology used for understanding and analysing the important safety culture issues at each nuclear power site is then described. This section begins with a summary of the processes used in the information gathering and data analysis stage. The six components of the Management Workshops conducted at each site are then described. These workshops used a series of presentations, interactive events and group exercises to: (a) provide feedback to site managers on the safety culture and safety leadership issues identified at their site, and (b) stimulate further safety thinking and provide 'take-away' information and leadership strategies that could be applied to promote safety culture improvements. Section 3, project Findings, contains the main observations and output from the project. These include: - a brief overview of aspects of the local industry operating context that impinge on safety culture; - a summary of strengths or positive attributes observed within the safety culture of the Swedish nuclear industry; - a set of identified opportunities for further improvement; - the aggregated

Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Katanishi, Shoji; Kunitomi, Kazuhiko

2003-01-01

Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Project of a binary breeder reactor and its inherent safety

International Nuclear Information System (INIS)

Nascimento, J.A. do; Dias, A.F.; Ishiguro, Y.

1983-01-01

A core layout for the binary breeder reactor (BBR) is developed based on the results of preliminary burnup calculations. The apparent breeding ratio, in the U 233 /Th fueled inner core, is low due to the accumulation of Pa-233 in the first few months of operation. The loss of reactivity during this time is around 3%. The BBR requires more reactivity control than Pu/U-fueled LMFBRs and the core layout developed has 19 control rod assemblies in the inner core. Three aspects related to the inherent safety of the Binary Breeder Reactor have been studied: the radial distribution of the sodium-void reactivity zone-wise Doppler reactivity and the fractions of delayed neutrons. The results show excellent characteristics for the BRB safety. (Author) [pt

Project of a binary breeder reactor and its inherent safety

International Nuclear Information System (INIS)

Nascimento, J.A. do; Dias, A.F.; Ishiguro, Y.

1983-01-01

A core layout for the binary breeder reactor (BBR) is developed based on the results of preliminary burnup calculations. In the U 233 /TH fueled inner core, the apparent breeding ratio is low due to the accumulation of Pa-233 in the first few months of operation. The loss of reactivity during this time is approximatelly 3%. The BBR requires more reactivity control than Pu/U-fueled LMFBRs and the core layout developed has 19 control rod assemblies in the inner core. Three aspects related to the inherent safety of the BBR have been studied: radial distribution of the sodium-void reactivity, zone-wise Doppler reactivity, and the delayed neutron fractions. Results show excellent safety characteristics of the BBR. (Author) [pt

Purification, crystallization and preliminary X-ray diffraction studies of parakeet (Psittacula krameri) haemoglobin

International Nuclear Information System (INIS)

Jaimohan, S. M.; Naresh, M. D.; Arumugam, V.; Mandal, A. B.

2009-01-01

Parakeet (Psittacula krameri) haemoglobin has been purified and crystallized under low salt buffered conditions. Preliminary analysis of the crystal that belonged to monoclinic system (C2) is reported. Birds often show efficient oxygen management in order to meet the special demands of their metabolism. However, the structural studies of avian haemoglobins (Hbs) are inadequate for complete understanding of the mechanism involved. Towards this end, purification, crystallization and preliminary X-ray diffraction studies have been carried out for parakeet Hb. Parakeet Hb was crystallized as the met form in low-salt buffered conditions after extracting haemoglobin from crude blood by microcentrifugation and purifying the sample by column chromatography. Good-quality crystals were grown from 10% PEG 3350 and a crystal diffracted to about 2.8 Å resolution. Preliminary diffraction data showed that the Hb crystal belonged to the monoclinic system (space group C2), with unit-cell parameters a = 110.68, b = 64.27, c = 56.40 Å, β = 109.35°. Matthews volume analysis indicated that the crystals contained a half-tetramer in the asymmetric unit

On the safety of conceptual fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Okrent, D.; Badham, V.; Caspi, S.; Chan, C.K.; Ferrell, W.J.; Frederking, T.H.K.; Grzesik, J.; Lee, J.Y.; McKone, T.E.; Pomraning, G.C.; Ullman, A.Z.; Ting, T.D.; Kim, Y.I.

1979-01-01

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium. As a result of these studies, it appears that highly reliable and even redundent decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered. (Auth.)

CFD Analysis of the Safety Injection Tank and Fluidic Device

Energy Technology Data Exchange (ETDEWEB)

Cho, Jai Oan; Nietiadi, Yohanes Setiawan; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Addad, Yacine [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-05-15

One of the most important components in the ECCS is the safety injection tank (SIT). Inside the SIT, a fluidic device is installed, which passively controls the mass flow of the safety injection and eliminates the need for low pressure safety injection pumps. As more passive safety mechanisms are being pursued, it has become more important to understand flow structure and the loss mechanism within the fluidic device. Current computational fluid dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study proposes to find a more exact result using CFD and more realistic modeling to predict the performance during accident scenarios more accurately. The safety injection tank with fluidic device was analyzed thoroughly using CFD. The preliminary calculation used 60,000 meshes for the initial test calculation. The results fit the experimental results surprisingly despite its coarse grid. Nonetheless, the mesh resolution was increased to capture the vortex in the fluidic device precisely. Once a detailed CFD computation is finished, a small-scale experiment will be conducted for the given conditions. Using the experimental results and the CFD model, physical models can be improved to fit the results more accurately.

Safety and skin delayed-type hypersensitivity response in vervet monkeys immunized with Leishmania donovani sonicate antigen delivered with adjuvants

OpenAIRE

Mutiso,Joshua M.; Macharia,John C.; Taracha,Evans; Wafula,Kellern; Rikoi,Hitler; Gicheru,Michael M.

2012-01-01

In this study, we report on the safety and skin delayed-type hypersensitivity (DTH), responses of the Leishmania donovani whole cell sonicate antigen delivered in conjunction with alum-BCG (AlBCG), Montanide ISA 720 (MISA) or Monophosphoryl lipid A (MPLA) in groups of vervet monkeys. Following three intradermal injections of the inoculums on days 0, 28 and 42, safety and DTH responses were assessed. Preliminary tumor necrosis factor alpha (TNF-α) and interferon gamma (IFN-γ) levels ...

Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

Probabilistic studies for a safety assurance program

International Nuclear Information System (INIS)

Iyer, S.S.; Davis, J.F.

1985-01-01

The adequate supply of energy is always a matter of concern for any country. Nuclear power has played, and will continue to play an important role in supplying this energy. However, safety in nuclear power production is a fundamental prerequisite in fulfilling this role. This paper outlines a program to ensure safe operation of a nuclear power plant utilizing the Probabilistic Safety Studies

Pediatric regional anesthesia: what is the current safety record?

Science.gov (United States)

Polaner, David M; Drescher, Jessica

2011-07-01

The use of regional anesthetics, whether as adjuncts, primary anesthetics or postoperative analgesia, is increasingly common in pediatric practice. Data on safety remain limited because of the paucity of very large-scale prospective studies that are necessary to detect low incidence events, although several studies either have been published or have reported preliminary results. This paper will review the data on complications and risk in pediatric regional anesthesia. Information currently available suggests that regional blockade, when performed properly, carries a very low risk of morbidity and mortality in appropriately selected infants and children. © 2010 Blackwell Publishing Ltd.

Activities on safety culture study. Study status in public and private sectors

International Nuclear Information System (INIS)

Makino, Maomi; Takano, Kenichi

2004-01-01

Around after entering in the 21st century, organizational accidents had occurred in Japan at various industries including nuclear industry, which were caused directly by unsafe action, human error and illegal conduct of personnel but there were problems in safety culture of organization such as slow retreat of safety system stimulated by management, schedule control and procedure management becoming a dead letter, lack of safety education, and workplace climate of schedule priority. This article referred to organizational factors common to many severe accidents and introduced safety culture study in public and private sectors to overcome those factors. Safety Culture Evaluation Support Tool (SCEST) was developed for self-evaluation of safety culture of organization as well as Organizational Reliability model (OR model) for analysis of initiation and propagation process of risk event. Safety diagnosis system and feedback type risk assessment system for promoting safe organizational climate and culture were also developed. (T. Tanaka)

Fire behaviour - A preliminary study. | W.S.W. | African Journal of ...

African Journals Online (AJOL)

Fire behaviour - A preliminary study. ... be taken cognisance of in any future research on fire in relation to vegetation. Keywords: behaviour; botany; environmental conditions; fire; fire behaviour; fire ecology; fires; grass; grasses; management; rate of spread; recovery; south africa; vegetation; veld; veld management; yield ...

Biocontamination Control for Spacesuit Garments - A Preliminary Study

Science.gov (United States)

Rhodes, Richard A.; Orndoff, Evelyne; Korona, F. Adam; Poritz, Darwin; Smith, Jelanie; Wong, Wing

2011-01-01

This paper outlines a preliminary study that was conducted to review, test, and improve on current space suit biocontamination control. Biocontamination from crew members can cause space suit damage and objectionable odors and lead to crew member health hazards. An understanding of the level of biocontamination is necessary to mitigate its effects. A series of tests were conducted with the intent of evaluating current suit materials, ground and on-orbit disinfectants, and potential commercial off-the-shelf antimicrobial materials. Included in this paper is a discussion of the test methodology, results, and analysis method.

Safety evaluation report related to the preliminary design of the Standard Nuclear Steam Supply Reference System, RESAR SP/90 (Docket No. STN 50-601)

International Nuclear Information System (INIS)

1991-04-01

On October 24, 1983, the Westinghouse Electric Corporation tendered its application for a preliminary design approval of the advanced pressurized-water reactor design for the SP/90 reactor. The Westinghouse Reference Safety Analysis Report (RESAR SP/90, Docket No. STN 50-601), describing the design of the facility, was submitted from October 24, 1983 through March 9, 1987. Staff of the US Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, has prepared this safety evaluation report of the RESAR SP/90 on the basis of its review. Because of the stage of the design, there are open issues that have not been resolved. These issues are discussed in detail throughout this report, and a summary is provided in Section 1.6 of this report. The applicant will be required to address these and any additional such concerns that may be raised during the course of the staff's review of advanced light-water reactors in support of a final design approval application. This report shall not constitute a commitment to issue a permit or license or in any way affect the authority of the Commission, its adjudicatory boards, and other presiding officers in any proceeding under Subpart G of Title 10 of the Code of Federal Regulations, Part 2

[European Union Network for Patient Safety and Quality of Care (PASQ). Development and preliminary results in Europe and in the Spanish National Health System].

Science.gov (United States)

Agra-Varela, Y; Fernández-Maíllo, M; Rivera-Ariza, S; Sáiz-Martínez-Acitorez, I; Casal-Gómez, J; Palanca-Sánchez, I; Bacou, J

2015-01-01

The joint action, European Union Network for Patient Safety and Quality of Care: PaSQ, aims to promote patient safety (PS) in the European Union (EU) and to facilitate the exchange of experiences among Member States (MS) and stakeholders on issues related to quality of care, PS, and patient involvement. The development and preliminary results are presented here, especially as regards the Spanish National Health System (SNHS). PaSQ is developed through 7 work packages, primarily aimed at sharing good practices (GP), which were identified using specific questionnaires and selected by means of explicit criteria, as well as to implement safe clinical practices (SCP) of proven effectiveness and agreed among MS. A total of 482 GP (39% provided by Spanish professionals) were identified. The 34 events organised in the EU, 11 including Spanish participation, facilitate sharing these practices. A total of 194 Health Care centres (49% in Spain) are implementing SCP (hand hygiene, safe surgery, medication reconciliation, and paediatric early warning scores) ACHIEVEMENTS AND FUTURE PERSPECTIVES: PaSQ is making it possible to strengthen collaboration between organizations and professionals at EU and SNHS level regarding PS and quality of care. Copyright © 2015 SECA. Published by Elsevier Espana. All rights reserved.

Protective Alternatives of SMR against Extreme Threat Scenario – A Preliminary Risk Analysis

International Nuclear Information System (INIS)

Shohet, I.M.; Ornai, D.; Gal, E.; Ronen, Y.; Vidra, M.

2014-01-01

The article presents a preliminary risk analysis of the main features in NPP (Nuclear Power Plant) that includes SMR - Small and Modular Reactors, given an extreme threat scenario. A review of the structure and systems of the SMR is followed by systematic definitions and analysis of the threat scenario to which a preliminary risk analysis was carried out. The article outlines the basic events caused by the referred threat scenario, which had led to possible failure mechanisms according to FTA (Fault-Tree-Analysis),critical protective circuits, and todetecting critical topics for the protection and safety of the reactor

Preliminary design for hot dirty-gas control-valve test facility. Final report

Energy Technology Data Exchange (ETDEWEB)

1980-01-01

This report presents the results of a preliminary design and cost estimating effort for a facility for the testing of control valves in Hot Dirty Gas (HDGCV) service. This design was performed by Mittelhauser Corporation for the United States Department of Energy's Morgantown Energy Technology Center (METC). The objective of this effort was to provide METC with a feasible preliminary design for a test facility which could be used to evaluate valve designs under simulated service conditions and provide a technology data base for DOE and industry. In addition to the actual preliminary design of the test facility, final design/construction/operating schedules and a facility cost estimate were prepared to provide METC sufficient information with which to evaluate this design. The bases, assumptions, and limitations of this study effort are given. The tasks carried out were as follows: METC Facility Review, Environmental Control Study, Gas Generation Study, Metallurgy Review, Safety Review, Facility Process Design, Facility Conceptual Layout, Instrumentation Design, Cost Estimates, and Schedules. The report provides information regarding the methods of approach used in the various tasks involved in the completion of this study. Section 5.0 of this report presents the results of the study effort. The results obtained from the above-defined tasks are described briefly. The turnkey cost of the test facility is estimated to be $9,774,700 in fourth quarter 1979 dollars, and the annual operating cost is estimated to be $960,000 plus utilities costs which are not included because unit costs per utility were not available from METC.

Preliminary standard review guide for Environmental Restoration/Decontamination and Decommissioning safety analyses

International Nuclear Information System (INIS)

Ellingson, D.R.

1993-06-01

The review guide is based on the shared experiences, approaches, and philosophies of the Environmental Restoration/Decontamination and Decommissioning (ER/D ampersand D) subgroup members. It is presented in the form of a review guide to maximize the benefit to both the safety analyses practitioner and reviewer. The guide focuses on those challenges that tend to be unique to ER/D ampersand D cleanup activities. Some of these experiences, approaches, and philosophies may find application or be beneficial to a broader spectrum of activities such as terminal cleanout or even new operations. Challenges unique to ER/D ampersand D activities include (1) consent agreements requiring activity startup on designated dates; (2) the increased uncertainty of specific hazards; and (3) the highly variable activities covered under the broad category of ER/D ampersand D. These unique challenges are in addition to the challenges encountered in all activities; e.g., new and changing requirements and multiple interpretations. The experiences in approaches, methods, and solutions to the challenges are documented from the practitioner and reviewer's perspective, thereby providing the viewpoints on why a direction was taken and the concerns expressed. Site cleanup consent agreements with predetermined dates for restoration activity startup add the dimension of imposed punitive actions for failure to meet the date. Approval of the safety analysis is a prerequisite to startup. Actions that increase expediency are (1) assuring activity safety; (2) documenting that assurance; and (3) acquiring the necessary approvals. These actions increase the timeliness of startup and decrease the potential for punitive action. Improvement in expediency has been achieved by using safety analysis techniques to provide input to the line management decision process rather than as a review of line management decisions. Expediency is also improved by sharing the safety input and resultant decisions with

Treatment of Retinopathy of Prematurity with topical ketorolac tromethamine: a preliminary study

Directory of Open Access Journals (Sweden)

Cafferata Maria

2004-08-01

Full Text Available Abstract Background Retinopathy of Prematurity (ROP is a common retinal neovascular disorder of premature infants. It is of variable severity, usually heals with mild or no sequelae, but may progress to blindness from retinal detachments or severe retinal scar formation. This is a preliminary report of the effectiveness and safety of a new and original use of topical ketorolac in preterm newborn to prevent the progression of ROP to the more severe forms of this disease. Methods From January 2001 to December 2002, all fifty nine preterm newborns with birthweight less than 1250 grams or gestational age less than 30 weeks of gestational age admitted to neonatal intensive care were eligible for treatment with topical ketorolac (0.25 milligrams every 8 hours in each eye. The historical comparison group included all 53 preterm newborns, with the same inclusion criteria, admitted between January 1999 and December 2000. Results Groups were comparable in terms of weight distribution, Apgar score at 5 minutes, incidence of sepsis, intraventricular hemorrhage and necrotizing enterocolitis. The duration of oxygen therapy was significantly longer in the control group. In the ketorolac group, among 43 children that were alive at discharge, one (2.3% developed threshold ROP and cryotherapy was necessary. In the comparison group 35 children survived, and six child (17% needed cryotherapy (Relative Risk 0.14, 95%CI 0.00 to 0.80, p = 0.041. Adjusting by duration of oxygen therapy did not significantly change these results. Adverse effects attributable to ketorolac were not detected. Conclusions This preliminary report suggests that ketorolac in the form of an ophthalmic solution can reduce the risk of developing severe ROP in very preterm newborns, without producing significant adverse side effects. These results, although promising, should be interpreted with caution because of the weakness of the study design. This is an inexpensive and simple intervention that

Thermal safety analysis of a dry storage cask for the Korean standard spent fuel - 16159

International Nuclear Information System (INIS)

Cha, Jeonghun; Kim, S.N.; Choi, K.W.

2009-01-01

A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed. (authors)

Standard model for the safety analysis report of nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

1980-02-01

This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization

The safety analysis of realization of the stabilization of beams B1/B2 supports project

International Nuclear Information System (INIS)

Aleshin, A.M.; Batij, V.G.; Glukhen'kij, V.N.; Kozoriz, V.I.; Klyuchnikov, A.A.; Kochnev, N.A.; Pavlovskij, L.I.; Rubezhanskij, Yu.I.; Sidorenko, N.V.; Stoyanov, A.I.; Shcherbin, V.N.

2000-01-01

The results of the analysis of radiation safety executed at preparing for realization of the stabilization of beams B1/B2 support are given. The comparison of results of the preliminary analysis of safety with the data received during realization of works is carried out

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

International Nuclear Information System (INIS)

Lowe, Andrew; Hayward, Brent

2006-08-01

This report documents a study with the objective of enhancing safety culture in the Swedish nuclear power industry. A primary objective of this study was to ensure that the latest thinking on human factors principles was being recognised and applied by nuclear power operators as a means of ensuring optimal safety performance. The initial phase of the project was conducted as a pilot study, involving the senior management group at one Swedish nuclear power-producing site. The pilot study enabled the project methodology to be validated after which it was repeated at other Swedish nuclear power industry sites, providing a broad-ranging analysis of opportunities across the industry to enhance safety culture. The introduction to this report contains an overview of safety culture, explains the background to the project and sets out the project rationale and objectives. The methodology used for understanding and analysing the important safety culture issues at each nuclear power site is then described. This section begins with a summary of the processes used in the information gathering and data analysis stage. The six components of the Management Workshops conducted at each site are then described. These workshops used a series of presentations, interactive events and group exercises to: (a) provide feedback to site managers on the safety culture and safety leadership issues identified at their site, and (b) stimulate further safety thinking and provide 'take-away' information and leadership strategies that could be applied to promote safety culture improvements. Section 3, project Findings, contains the main observations and output from the project. These include: - a brief overview of aspects of the local industry operating context that impinge on safety culture; - a summary of strengths or positive attributes observed within the safety culture of the Swedish nuclear industry; - a set of identified opportunities for further improvement; - the aggregated results of the

Nuclear Safety in Central and Eastern Europe

International Nuclear Information System (INIS)

2001-04-01

Nuclear safety is one of the critical issues with respect to the enlargement of the European Union towards the countries of Central and Eastern Europe. In the context of the enlargement process, the European Commission overall strategy on nuclear safety matters has been to bring the general standard of nuclear safety in the pre-accession countries up to a level that would be comparable to the safety levels in the countries of the European Union. In this context, the primary objective of the project was to develop a common format and general guidance for the evaluation of the current nuclear safety status in countries that operate commercial nuclear power plants. Therefore, one of the project team first undertakings was to develop an approach that would allow for a consistent and comprehensive overview of the nuclear safety status in the CEEC, enabling an equal treatment of the countries to be evaluated. Such an approach, which did not exist, should also ensure identification of the most important safety issues of the individual nuclear power plants. The efforts resulted in the development of the ''Performance Evaluation Guide'', which focuses on important nuclear safety issues such as plant design and operation, the practice of performing safety assessments, and nuclear legislation and regulation, in particular the role of the national regulatory body. Another important aspect of the project was the validation of the Performance Evaluation Guide (PEG) by performing a preliminary evaluation of nuclear safety in the CEEC, namely in Bulgaria, Czech Republic, Hungary, Lithuania, Romania, Slovak Republic, and Slovenia. The nuclear safety evaluation of each country was performed as a desktop exercise, using solely available documents that had been prepared by various Western institutions and the countries themselves. Therefore, the evaluation is only of a preliminary nature. The project did not intend to re-assess nuclear safety, but to focus on a comprehensive summary

Nanotechnology Safety Self-Study

Energy Technology Data Exchange (ETDEWEB)

Grogin, Phillip W. [Los Alamos National Laboratory

2016-03-29

Nanoparticles are near-atomic scale structures between 1 and 100 nanometers (one billionth of a meter). Engineered nanoparticles are intentionally created and are used in research and development at Sandia National Laboratories (SNL) and Los Alamos National Laboratory (LANL). This course, Nanotechnology Safety Self-Study, presents an overview of the hazards, controls, and uncertainties associated with the use of unbound engineered nanoscale particles (UNP) in a laboratory environment.

78 FR 64504 - Safety and Occupational Health Study Section (SOHSS), National Institute for Occupational Safety...

Science.gov (United States)

2013-10-29

... DEPARTMENT OF HEALTH AND HUMAN SERVICES Centers for Disease Control and Prevention Safety and Occupational Health Study Section (SOHSS), National Institute for Occupational Safety and Health (NIOSH or..., Number 177, Pages 56235-56236. Contact Person for More Information: Price Connor, Ph.D., NIOSH Health...

A preliminary case series evaluating the safety and immediate to short-term clinical benefits of joint mobilization in hemophilic arthritis of the lower limb.

Science.gov (United States)

Scaddan, Emma; Rowell, John; O'Leary, Shaun

2017-09-01

Arthritis resulting from recurrent intra-articular bleeding in individuals with hemophilia can be severely debilitating due to joint pain and stiffness with subsequent loss of mobility and function. Very limited studies have investigated the potential benefits of joint mobilization for this condition. This case series is a preliminary investigation of safety, as well as immediate and short-term clinical benefits, associated with gentle knee and ankle joint mobilization in people with hemophilic arthropathy. A single intervention of joint mobilization was applied to the affected knees and/or ankles of 16 individuals with severe or moderate hemophilia within a public hospital setting. Adverse events, as well as immediate (pain-free passive joint range, Timed Up and Go Test with maximum pain numerical rating scale) and short-term (Lower Extremity Functional Scale) effects of the intervention were evaluated with a repeated measures ANOVA. There were no adverse events. An immediate significant increase was observed in pain-free passive ankle joint range of motion ( p  < 0.05) following the joint mobilization intervention. The findings of this case series suggest that gentle joint mobilization techniques may be safely considered as part of a multimodal management approach for hemophilic arthropathy.

Preliminary nuclear decommissioning cost study

International Nuclear Information System (INIS)

Sissingh, R.A.P.

1981-04-01

The decommissioning of a nuclear power plant may involve one or more of three possible options: storage with surveillance (SWS), restricted site release (RSR), and unrestricted site use(USU). This preliminary study concentrates on the logistical, technical and cost aspects of decommissioning a multi-unit CANDU generating station using Pickering GS as the reference design. The procedure chosen for evaluation is: i) removal of the fuel and heavy water followed by decontamination prior to placing the station in SWS for thiry years; ii) complete dismantlement to achieve a USU state. The combination of SWS and USU with an interim period of surveillance allows for radioactive decay and hence less occupational exposure in achieving USU. The study excludes the conventional side of the station, assumes waste disposal repositories are available 1600 km away from the station, and uses only presently available technologies. The dismantlement of all systems except the reactor core can be accomplished using Ontario Hydro's current operating, maintenance and construction procedures. The total decommissioning period is spread out over approximately 40 years, with major activities concentrated in the first and last five years. The estimated dose would be approximately 1800 rem. Overall Pickering GS A costs would be $162,000,000 (1980 Canadian dollars)

Canadian Consumer Food Safety Practices and Knowledge: Foodbook Study.

Science.gov (United States)

Murray, Regan; Glass-Kaastra, Shiona; Gardhouse, Christine; Marshall, Barbara; Ciampa, Nadia; Franklin, Kristyn; Hurst, Matt; Thomas, M Kate; Nesbitt, Andrea

2017-10-01

Understanding consumers' food safety practices and knowledge supports food safety education for the prevention of foodborne illness. The objective of this study was to describe Canadian consumer food safety practices and knowledge. This study identifies demographic groups for targeted food safety education messaging and establishes a baseline measurement to assess the effectiveness of food safety interventions over time. Questions regarding consumer food safety practices and knowledge were included in a population-based telephone survey, Foodbook, conducted from November 2014 to March 2015. The results were analyzed nationally by age group and by gender. The results showed that approximately 90% of Canadians reported taking the recommended cleaning and separating precautions when handling raw meat to prevent foodborne illness. Only 29% of respondents reported using a food thermometer when cooking any meat, and even fewer (12%) reported using a food thermometer for small cuts of meat such as chicken pieces. The majority (>80%) of Canadians were aware of the foodborne illness risks related to chicken and hamburger, but fewer (poultry.

Continuous improvement of the MHTGR safety and competitive performance

International Nuclear Information System (INIS)

Eichenberg, T.W.; Etzel, K.T.; Mascaro, L.L.; Rucker, R.A.

1992-05-01

An increase in reactor module power from 350 to 450 MW(t) would markedly improve the economics of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The higher power level was recommended as the result of an in-depth cost reduction study undertaken to compete with the declining price of fossil fuel. The safety assessment confirms that the high level of safety, which relies on inherent characteristics and passive features, is maintained at the elevated power level. Preliminary systems, nuclear, and safety performance results are discussed for the recommended 450 MW(t) design. Optimization of plant parameters and design modifications accommodated the operation of the steam generator and circulator at the higher power level. Events in which forced cooling is lost, designated as conduction cooldowns are described in detail. For the depressurized conduction cooldown, without full helium inventory, peak fuel temperatures are significantly lowered. A more negative temperature coefficient of reactivity was achieved while maintaining an adequate fuel cycle and reactivity control. Continual improvement of the MHTGR delivers competitive performance without relinquishing the high safety margins demanded of the next generation of power plants