Qualification issues associated with the use of advanced instrumentation and control systems hardware in nuclear power plants

International Nuclear Information System (INIS)

Korsah, K.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and Tiber-optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light water reactor. The template was then used to address reliability issues for microprocessor-based protection systems. Standards (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems were also identified. This approach addresses in part issues raised in Nuclear Regulatory Commission policy document SECY-91-292. which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards, as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

International Nuclear Information System (INIS)

Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

2003-01-01

The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

Laser light scattering instrument advanced technology development

Science.gov (United States)

Wallace, J. F.

1993-01-01

The objective of this advanced technology development (ATD) project has been to provide sturdy, miniaturized laser light scattering (LLS) instrumentation for use in microgravity experiments. To do this, we assessed user requirements, explored the capabilities of existing and prospective laser light scattering hardware, and both coordinated and participated in the hardware and software advances needed for a flight hardware instrument. We have successfully breadboarded and evaluated an engineering version of a single-angle glove-box instrument which uses solid state detectors and lasers, along with fiber optics, for beam delivery and detection. Additionally, we have provided the specifications and written verification procedures necessary for procuring a miniature multi-angle LLS instrument which will be used by the flight hardware project which resulted from this work and from this project's interaction with the laser light scattering community.

Advanced instrumentation and teleoperation

International Nuclear Information System (INIS)

Decreton, M.

1998-01-01

SCK-CEN's advanced instrumentation and teleoperation project aims at evaluating the potential of a telerobotic approach in a nuclear environment and, in particular, the use of remote-perception systems. Main achievements in 1997 in the areas of R and D on radiation tolerance for remote sensing, optical fibres and optical-fibre sensors, and computer-aided teleoperation are reported

New technology in nuclear power plant instrumentation and control

International Nuclear Information System (INIS)

Anon.

1985-01-01

The primary topic of this book is what can be done to improve nuclear power plant operation safety and the economic benefits that can be gained with the utilization of advance instrumentation and control technology. Other topics discussed are the industry's reluctance to accept new designs determining cost effective improvements, and difficulties in meeting regulatory standards with new technology control. The subjects will be useful when considering the area of instrumentation and control for enhancing plant operation and safety. Contents: Advanced Instrumention, Plant Control and Monitoring, Plant Diagnostics and Failure Detection, Human Factors Considerations in Instrumentation and Control, NRC and Industry Perspective on Advanced Instrumentation and Control

Qualification issues associated with the use of advanced instrumentation and control systems hardware in nuclear power plants

International Nuclear Information System (INIS)

Korsah, K.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber-optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light water reactor. The template was then used to address reliability issues for microprocessor-based protection systems. Standards (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems were also identified. This approach addresses in part issues raised in Nuclear Regulatory Commission policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are open-quotes being developed without consensus standards, as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.close quotes

Reactor safety instrumentation of Paks NPP (experience and perspective)

International Nuclear Information System (INIS)

Elo, S.; Hamar, K.

1993-01-01

The majority of the existing control and protection systems in nuclear power plants use old analog technology and design philosophy. Maintenance and the procurement of spare parts is becoming increasingly difficult. In general there is an age degradation concern. Aging degradation in nuclear power plants must be effectively managed to avoid a loss of vital safety function, shutdown of the station, a reduced power generation, or any failure leading to expensive repair. Even with the best efforts in developing reliable and long life instrumentation and control systems for nuclear power plants it is expected that these systems for most plants will require replacements during the life of the plants. The instrumentation and control system of the nuclear power plants designed during the 70's and constructed in the 80's went out-of-date since nuclear safety is not a static concept and the digital computer technology has undergone rapid improvements during the 70's and 80's. Simultaneously the operation and the maintenance of the I ampersand C system of those plants described above becomes more and more difficult and expensive. In this context the pure quality of the former Soviet designed process instrumentation system increases the needs of upgrading this system. The author reviews the main design characteristics of the reactor safety instrumentation of the Paks NPP. Further he attempts to convey the perspective on upgrading the reactor safety instrumentation as seen by the HAEC and its Nuclear Safety Inspectorate

Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

Evaluation of robotically controlled advanced endoscopic instruments

NARCIS (Netherlands)

Reilink, Rob; Kappers, Astrid M.L.; Stramigioli, Stefano; Misra, Sarthak

Background Advanced flexible endoscopes and instruments with multiple degrees of freedom enable physicians to perform challenging procedures such as the removal of large sections of mucosal tissue. However, these advanced endoscopes are difficult to control and require several physicians to

Conceptual design of safety instrumentation for PFBR

International Nuclear Information System (INIS)

Muralikrishna, G.; Seshadri, U.; Raghavan, K.

1996-01-01

Instrumentation systems enable monitoring of the process which in turn enables control and shutdown of the process as per the requirements. Safety Instrumentation due to its vital importance has a stringent role and this needs to be designed methodically. This paper presents the details of the conceptual design for PFBR. (author). 4 figs, 3 tabs

Standards for radiation protection instrumentation: design of safety standards and testing procedures

International Nuclear Information System (INIS)

Meissner, Frank

2008-01-01

This paper describes by means of examples the role of safety standards for radiation protection and the testing and qualification procedures. The development and qualification of radiation protection instrumentation is a significant part of the work of TUV NORD SysTec, an independent expert organisation in Germany. The German Nuclear Safety Standards Commission (KTA) establishes regulations in the field of nuclear safety. The examples presented may be of importance for governments and nuclear safety authorities, for nuclear operators and for manufacturers worldwide. They demonstrate the advantage of standards in the design of radiation protection instrumentation for new power plants, in the upgrade of existing instrumentation to nuclear safety standards or in the application of safety standards to newly developed equipment. Furthermore, they show how authorities may proceed when safety standards for radiation protection instrumentation are not yet established or require actualization. (author)

Advanced optical instruments technology

Science.gov (United States)

Shao, Mike; Chrisp, Michael; Cheng, Li-Jen; Eng, Sverre; Glavich, Thomas; Goad, Larry; Jones, Bill; Kaarat, Philip; Nein, Max; Robinson, William

1992-08-01

The science objectives for proposed NASA missions for the next decades push the state of the art in sensitivity and spatial resolution over a wide range of wavelengths, including the x-ray to the submillimeter. While some of the proposed missions are larger and more sensitive versions of familiar concepts, such as the next generation space telescope, others use concepts, common on the Earth, but new to space, such as optical interferometry, in order to provide spatial resolutions impossible with other concepts. However, despite their architecture, the performance of all of the proposed missions depends critically on the back-end instruments that process the collected energy to produce scientifically interesting outputs. The Advanced Optical Instruments Technology panel was chartered with defining technology development plans that would best improve optical instrument performance for future astrophysics missions. At this workshop the optical instrument was defined as the set of optical components that reimage the light from the telescope onto the detectors to provide information about the spatial, spectral, and polarization properties of the light. This definition was used to distinguish the optical instrument technology issues from those associated with the telescope, which were covered by a separate panel. The panel identified several areas for optical component technology development: diffraction gratings; tunable filters; interferometric beam combiners; optical materials; and fiber optics. The panel also determined that stray light suppression instruments, such as coronagraphs and nulling interferometers, were in need of general development to support future astrophysics needs.

The Effects of Advanced 'Glass Cockpit' Displayed Flight Instrumentation on In-flight Pilot Decision Making

Science.gov (United States)

Steigerwald, John

The Cognitive Continuum Theory (CCT) was first proposed 25 years ago to explain the relationship between intuition and analytical decision making processes. In order for aircraft pilots to make these analytical and intuitive decisions, they obtain information from various instruments within the cockpit of the aircraft. Advanced instrumentation is used to provide a broad array of information about the aircraft condition and flight situation to aid the flight crew in making effective decisions. The problem addressed is that advanced instrumentation has not improved the pilot decision making in modern aircraft. Because making a decision is dependent upon the information available, this experimental quantitative study sought to determine how well pilots organize and interpret information obtained from various cockpit instrumentation displays when under time pressure. The population for this study was the students, flight instructors, and aviation faculty at the Middle Georgia State College School of Aviation campus in Eastman, Georgia. The sample was comprised of two groups of 90 individuals (45 in each group) in various stages of pilot licensure from student pilot to airline transport pilot (ATP). The ages ranged from 18 to 55 years old. There was a statistically significant relationship at the p safety of flight.

Comparing NICU teamwork and safety climate across two commonly used survey instruments.

Science.gov (United States)

Profit, Jochen; Lee, Henry C; Sharek, Paul J; Kan, Peggy; Nisbet, Courtney C; Thomas, Eric J; Etchegaray, Jason M; Sexton, Bryan

2016-12-01

Measurement and our understanding of safety culture are still evolving. The objectives of this study were to assess variation in safety and teamwork climate and in the neonatal intensive care unit (NICU) setting, and compare measurement of safety culture scales using two different instruments (Safety Attitudes Questionnaire (SAQ) and Hospital Survey on Patient Safety Culture (HSOPSC)). Cross-sectional survey study of a voluntary sample of 2073 (response rate 62.9%) health professionals in 44 NICUs. To compare survey instruments, we used Spearman's rank correlation coefficients. We also compared similar scales and items across the instruments using t tests and changes in quartile-level performance. We found significant variation across NICUs in safety and teamwork climate scales of SAQ and HSOPSC (pteamwork scales (teamwork climate and teamwork within units) of the two instruments correlated strongly (safety r=0.72, pteamwork r=0.67, p<0.001). However, the means and per cent agreements for all scale scores and even seemingly similar item scores were significantly different. In addition, comparisons of scale score quartiles between the two instruments revealed that half of the NICUs fell into different quartiles when translating between the instruments. Large variation and opportunities for improvement in patient safety culture exist across NICUs. Important systematic differences exist between SAQ and HSOPSC such that these instruments should not be used interchangeably. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

Use of modern software - based instrumentation in safety critical systems

International Nuclear Information System (INIS)

Emmett, J.; Smith, B.

2005-01-01

Many Nuclear Power Plants are now ageing and in need of various degrees of refurbishment. Installed instrumentation usually uses out of date 'analogue' technology and is often no longer available in the market place. New technology instrumentation is generally un-qualified for nuclear use and specifically the new 'smart' technology contains 'firmware', (effectively 'soup' (Software of Uncertain Pedigree)) which must be assessed in accordance with relevant safety standards before it may be used in a safety application. Particular standards are IEC 61508 [1] and the British Energy (BE) PES (Programmable Electronic Systems) guidelines EPD/GEN/REP/0277/97. [2] This paper outlines a new instrument evaluation system, which has been developed in conjunction with the UK Nuclear Industry. The paper concludes with a discussion about on-line monitoring of Smart instrumentation in safety critical applications. (author)

Recent development of nuclear power in Japan and instrumentation and control system and control room equipment for advanced light water reactors

International Nuclear Information System (INIS)

Wakayama, N.

1992-01-01

This paper was provided for the 13th IAEA/IWG-NPPCI Meeting and aims to introduce an outline of recent development of nuclear power in Japan and some topics in the field of nuclear power plant control and instrumentation. Forty units of nuclear power plants are in operation in Japan and five units of BWRs and six PWRs are under construction. Construction of prototype FBR Monju have almost completed an construction of High-Temperature Engineering Test Reactor, HTTR, started in March 1991. In parallel of those, extensive effort has been carried out to develop the third generation LWRs which are called Advanced BWR (ABWR) and Advanced PWR (APWR). Two Advanced BWRs are under safety review for construction. Instrumentation and control system of these Advanced LWRs adopts integrated digital I and C system, optical multiplexing signal transmission, fault tolerant control systems and software logic for reactor protection and safety systems and enhances plant control performance and provides human-friendly operation and maintenance environments. Main control room of these Advanced LWRs, comprised with large display panels and advanced console, has special futures such as one-man sit-down operation, human friendly man-machine interface, high level automation in operation and maintenance. (author). 7 refs, 9 figs, 1 tab

Instrumentation

International Nuclear Information System (INIS)

Umminger, K.

2008-01-01

A proper measurement of the relevant single and two-phase flow parameters is the basis for the understanding of many complex thermal-hydraulic processes. Reliable instrumentation is therefore necessary for the interaction between analysis and experiment especially in the field of nuclear safety research where postulated accident scenarios have to be simulated in experimental facilities and predicted by complex computer code systems. The so-called conventional instrumentation for the measurement of e. g. pressures, temperatures, pressure differences and single phase flow velocities is still a solid basis for the investigation and interpretation of many phenomena and especially for the understanding of the overall system behavior. Measurement data from such instrumentation still serves in many cases as a database for thermal-hydraulic system codes. However some special instrumentation such as online concentration measurement for boric acid in the water phase or for non-condensibles in steam atmosphere as well as flow visualization techniques were further developed and successfully applied during the recent years. Concerning the modeling needs for advanced thermal-hydraulic codes, significant advances have been accomplished in the last few years in the local instrumentation technology for two-phase flow by the application of new sensor techniques, optical or beam methods and electronic technology. This paper will give insight into the current state of instrumentation technology for safety-related thermohydraulic experiments. Advantages and limitations of some measurement processes and systems will be indicated as well as trends and possibilities for further development. Aspects of instrumentation in operating reactors will also be mentioned.

Advanced nuclear reactor safety issues and research needs

International Nuclear Information System (INIS)

2002-01-01

On 18-20 February 2002, the OECD Nuclear Energy Agency (NEA) organised, with the co-sponsorship of the International Atomic Energy Agency (IAEA) and in collaboration with the European Commission (EC), a Workshop on Advanced Nuclear Reactor Safety Issues and Research Needs. Currently, advanced nuclear reactor projects range from the development of evolutionary and advanced light water reactor (LWR) designs to initial work to develop even further advanced designs which go beyond LWR technology (e.g. high-temperature gas-cooled reactors and liquid metal-cooled reactors). These advanced designs include a greater use of advanced technology and safety features than those employed in currently operating plants or approved designs. The objectives of the workshop were to: - facilitate early identification and resolution of safety issues by developing a consensus among participating countries on the identification of safety issues, the scope of research needed to address these issues and a potential approach to their resolution; - promote the preservation of knowledge and expertise on advanced reactor technology; - provide input to the Generation IV International Forum Technology Road-map. In addition, the workshop tried to link advancement of knowledge and understanding of advanced designs to the regulatory process, with emphasis on building public confidence. It also helped to document current views on advanced reactor safety and technology, thereby contributing to preserving knowledge and expertise before it is lost. (author)

Safety regulations concerning instrumentation and control systems for research reactors

International Nuclear Information System (INIS)

El-Shanshoury, A.I.

2009-01-01

A brief study on the safety and reliability issues related to instrumentation and control systems in nuclear reactor plants is performed. In response, technical and strategic issues are used to accomplish instrumentation and control systems safety. For technical issues there are ; systems aspects of digital I and C technology, software quality assurance, common-mode software, failure potential, safety and reliability assessment methods, and human factors and human machine interfaces. The strategic issues are the case-by-case licensing process and the adequacy of the technical infrastructure. The purpose of this work was to review the reliability of the safety systems related to these technical issues for research reactors

Instrumentation to Enhance Advanced Test Reactor Irradiations

Energy Technology Data Exchange (ETDEWEB)

J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

2009-09-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

Instrumentation to Enhance Advanced Test Reactor Irradiations

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Daw, J.E.; Taylor, S.C.

2009-01-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

Standardization and improvement of safety for radioisotope equipped instruments

International Nuclear Information System (INIS)

Sumi, Tetsuo

1980-01-01

The safety for radioisotope-equipped instruments is considered. The one is the safety for the source assembly. The radioisotopes employed for radioisotope-equipped instruments are sealed sources which are used in the state of being contained in the enclosures. Many of the enclosures are provided with shutter mechanism for the purpose of emitting radiation only during the period required. If the possible troubles that might lead to the accidents are sampled out of the results of field operation of radiation instruments, and the safety measures for source enclosures are considered in connection with these troubles, it is no exaggeration to say that the safety for source enclosures has been maintained by preventing the critical accidents by the management of users and the cooperation of manufactures though there were the chance for investigating the safety in the common field and the establishment of JIS Z 4614 standard. Another consideration is concerned with the measures to improve the safety. No accident in the past never guarantees no accident in the future. Accumulation of experience is most effective for those measures, and the more experiences the better. It may be most effective that the manufacturers disclose their experiences each other from the wide outlook overcoming the barrier of trade secret. Fortunately, such consciousness has risen since a few years ago, and the investigation group is doing the works in the Japan Radioisotope Association. On the other hand, the reasonable revision of the radiation injury prevention law is desired. (Wakatsuki, Y.)

Proceedings of national symposium on advanced instrumentation for nuclear research

International Nuclear Information System (INIS)

1993-01-01

The National Symposium on Advanced Instrumentation for Nuclear Research was held in Bombay during January 27-29, 1993 at BARC. Progress of modern nuclear research is closely related to the availability of state of the art instruments and systems. With the advancements in experimental techniques and sophisticated detector developments, the performance specifications have become more stringent. State of the art techniques and diverse applications of sophisticated nuclear instrumentation systems are discussed along with indigenous efforts to meet the specific instrumentation needs of research programs in nuclear sciences. Papers of relevance to nuclear science and technology are indexed separately. (original)

Development of advanced instrumentation for the Phebus FP project-preliminary studies

International Nuclear Information System (INIS)

Hampel, G.; Poss, G.

1989-10-01

The objective of the project was to examine advanced measuring methods for on-line determination of aerosol, thermohydraulic and hydrogen parameters for their usefulness in the French core melting programme PHEBUS FP. Advanced methods that are known from reactor safety programmes as well as novel measuring techniques as used in related fields such as fuel reprocessing and in non-nuclear process technology have been described and analysed with a view to their potential application in the planned tests. Possibly necessary further developments and modifications have been identified and approaches to solution have been outlined. Special emphasis has been put on the present state of radiation resistant optical fibre wave-guides, which might be integrated in several instruments. As regards the particularly important on-line aerosol measurements in PHEBUS FP, a combined system is recommended as the most promising further development. This novel system configuration comprises the mass monitor of a MESA dust monitoring system and one or several photometers of EIR design

Advanced Technologies and Instrumentation at the National Science Foundation

Science.gov (United States)

Kurczynski, Peter; Neff, James E.

2018-01-01

Over its more than thirty-year history, the Advanced Technologies and Instrumentation (ATI) program within the Division of Astronomical Sciences has provided grants to support the development and deployment of detectors and instrumentation for ground-based astronomy. This program has enabled scientific advances in diverse fields from solar physics to exoplanets to cosmology. ATI has provided instrumentation for both small and large observatories from radio through visible wavebands. It has played a role in the early development of major initiatives such as the Large Synoptic Survey Telescope. Technology development for astronomy unfolds over a longer period than the lifetime of a single grant. This review will consider ATI from an historical perspective to assess its impact on astronomy.

Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2015-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

Impact of state-of-the-art instrumentation on safety-related experimental studies proposed in containment studies facility (CSF)

International Nuclear Information System (INIS)

Gole, N.V.; Markandeya, S.G.; Subramaniam, K.; Ghosh, A.K.

2002-01-01

Full text: Conducting an experimental program for safety related studies for nuclear power plants (NPPs) is an extremely laborious and time-consuming task due to several reasons. Requirement for frequent replacements, testing and recalibration of a large number of instruments is one of them. Off-line analysis leading to identification of errors is another. A particular test may have to be abandoned based on such analysis. Following the rapid advances in instrumentation, a larger number of options are now available, which make experimentation easy. CSF is one of the upcoming facilities wherein deployment of state-of-the art became inevitable. This paper discusses in detail the design intent of instrumentation, the state-of-the-art instrumentation provisions made to fulfill it the overall impact of this on successful experimentation

Advanced Reactor Safety Program – Stakeholder Interaction and Feedback

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-08-01

In the Spring of 2013, the Idaho National Laboratory (INL) began discussions with industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling and simulation could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

Issues affecting advanced passive light-water reactor safety analysis

International Nuclear Information System (INIS)

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

Safety criteria for advanced HTGR concepts

International Nuclear Information System (INIS)

Kroeger, W.

1989-01-01

It is commonly agreed that advanced HTGR concepts must be licensable, which means that they must fulfil existing regulatory requirements. Furthermore, it is necessary to improve their public acceptance and they must even be suitable for urban sites. Therefore, they should be 'safer' than existing plants, which mainly means with respect to low-frequency or beyond-design severe accidents. Last but not least, the realization of advanced HTGR would be easier if commonly shared safety principles could be stated ensuring this further increased level of safety internationally. These qualitative statements need to be cast into quantitative guidelines which can be used as a rationale for safety evaluation. This paper tries to describe the status reached and to stimulate international activities. (author). 12 refs, 4 figs, 3 tabs

Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

International Nuclear Information System (INIS)

Holcomb, David Eugene

2015-01-01

Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate

Proceedings of the OECD/CSNI specialist meeting on advanced instrumentation and measurement techniques

Energy Technology Data Exchange (ETDEWEB)

Lehner, J [comp.

1998-09-01

In the last few years, tremendous advances in the local instrumentation technology for two-phase flow have been accomplished by the applications of new sensor techniques, optical or beam methods and electronic technology. The detailed measurements gave new insight to the true nature of local mechanisms of interfacial transfer between phases, interfacial structure and two-phase flow turbulent transfers. These new developments indicate that more accurate and reliable two-phase flow models can be obtained, if focused experiments are designed and performed by utilizing this advanced instrumentation. The purpose of this Specialist Meeting on Advanced Instrumentation and Measurement Techniques was to review the recent instrumentation developments and the relation between thermal-hydraulic codes and instrumentation capabilities. Four specific objectives were identified for this meeting: bring together international experts on instrumentation, experiments, and modeling; review recent developments in multiphase flow instrumentation; discuss the relation between modeling needs and instrumentation capabilities, and discuss future directions for instrumentation development, modeling, and experiments.

Proceedings of the OECD/CSNI specialist meeting on advanced instrumentation and measurement techniques

International Nuclear Information System (INIS)

Lehner, J.

1998-09-01

In the last few years, tremendous advances in the local instrumentation technology for two-phase flow have been accomplished by the applications of new sensor techniques, optical or beam methods and electronic technology. The detailed measurements gave new insight to the true nature of local mechanisms of interfacial transfer between phases, interfacial structure and two-phase flow turbulent transfers. These new developments indicate that more accurate and reliable two-phase flow models can be obtained, if focused experiments are designed and performed by utilizing this advanced instrumentation. The purpose of this Specialist Meeting on Advanced Instrumentation and Measurement Techniques was to review the recent instrumentation developments and the relation between thermal-hydraulic codes and instrumentation capabilities. Four specific objectives were identified for this meeting: bring together international experts on instrumentation, experiments, and modeling; review recent developments in multiphase flow instrumentation; discuss the relation between modeling needs and instrumentation capabilities, and discuss future directions for instrumentation development, modeling, and experiments

Evaluation of advanced two-phase flow instrumentation in SCTF Core-1

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Sudo, Yukio; Adachi, Hiromichi

1984-03-01

In the Slab Core Test Facility (SCTF) Core-I, advanced two-phase flow instruments have been provided by the USNRC to measure the thermohydraulic behavior in the primary system including pressure vessel during the end of blowdown, refill and reflood phases of a postulated loss-of-coolant accident in a pressurized water reactor. The advanced instruments are turbine meters, drag disks, γ-densitometers, spool pieces, liquid level detectors (LLD), fluid distribution grids (FDG), impedance probes (flag, prong and string probes), film probes, and video optical probes. This report presents evaluated results of the data from these instruments. Some instruments are quantitatively evaluated by comparing with the data from the conventional instruments or the other advanced instruments. Main conclusions are as follows: (1) The spool pieces and the γ-densitometers work well and provide satisfactory results; (2) Some of the turbine meters, the impedance probes and the film probes give partially reasonable results, but still more improvements are required; (3) Most of the LLDs, the FDGs, the impedance probes, and the film probes do not work well due to a hard cable corrosion, and (4) The video optical probes give clear image of the flow pattern. (author)

New trends in pile safety instrumentation; Les tendances nouvelles dans l'instrumentation de securite des piles

Energy Technology Data Exchange (ETDEWEB)

Furet, J.

1961-04-19

This report addresses the protection of nuclear piles against damages due to operation incidents. The author discusses the current trends in the philosophy of safety of atomic power piles, identifies the parameters which define safety systems, presents tests to be performed on safety chains, comments the relationship between safety and the decrease of the number of pile inadvertent shutdowns, discusses the issues of instrument failures and chain multiplicity, comments the possible improvement of the operation of elements which build up safety chains (design simplification, development of semiconductors, replacement of electromechanical relays by static relays), the role of safety logical computers and the development of automatics in pile safety, presents automatic control as a safety factor (example of automatic start-up), and finally comments the use of fuses.

Analysis instruments for the performance of Advanced Practice Nursing.

Science.gov (United States)

Sevilla-Guerra, Sonia; Zabalegui, Adelaida

2017-11-29

Advanced Practice Nursing has been a reality in the international context for several decades and recently new nursing profiles have been developed in Spain as well that follow this model. The consolidation of these advanced practice roles has also led to of the creation of tools that attempt to define and evaluate their functions. This study aims to identify and explore the existing instruments that enable the domains of Advanced Practice Nursing to be defined. A review of existing international questionnaires and instruments was undertaken, including an analysis of the design process, the domains/dimensions defined, the main results and an exploration of clinimetric properties. Seven studies were analysed but not all proved to be valid, stable or reliable tools. One included tool was able to differentiate between the functions of the general nurse and the advanced practice nurse by the level of activities undertaken within the five domains described. These tools are necessary to evaluate the scope of advanced practice in new nursing roles that correspond to other international models of competencies and practice domains. Copyright © 2017 Elsevier España, S.L.U. All rights reserved.

ALADIN - Advanced Laue Diffraction Instruments using Neutrons

International Nuclear Information System (INIS)

Lemee-Cailleau, M.H.; Ouladdiaf, B.; McIntyre, G.J.

2011-01-01

Laue diffraction techniques have proven to be very attractive to a broad user community interested in obtaining detailed structural information on very small single-crystal samples or needing data collection speeds comparable to those available with the powder diffraction technique. However our experience has clearly demonstrated the negative effect of up-stream monochromatic instruments on the quality of Laue data. In order to obtain Laue diffraction data with a statistical accuracy similar to that achieved on a monochromatic instrument (neutron or X-rays), the project ALADIN (for Advanced Laue Diffraction Instruments using Neutrons) aims to: -) construct a Laue-dedicated thermal neutron guide, with m=2 super-mirror coating, providing access to the desirable wavelength bandwidth; -) installation of one of the ILL Laue diffractometers (VIVALDI or CYCLOPS) on this new guide. (authors)

Advanced Instrumentation and control techniques for nuclear power plants

International Nuclear Information System (INIS)

Mori, Nobuyuki; Makino, Maomi; Naito, Norio

1992-01-01

Toshiba has been promoting the development of an advanced instrumentation and control system for nuclear power plants to fulfill the requirements for increased reliability, improved functionality and maintainability, and more competitive economic performance. This system integrates state-of-the-art technologies such as those for the latest man-machine interface, digital processing, optical multiplexing signal transmission, human engineering, and artificial intelligence. Such development has been systematically accomplished based on a schematic view of integrated digital control and instrumentation systems, and the development of whole systems has now been completed. This paper describes the purpose, design philosophy, and contents of newly developed systems, then considers the future trends of advanced man-machine systems. (author)

Safety issues on advanced fuel

International Nuclear Information System (INIS)

Gross, H.; Krebs, W.D.

1998-01-01

In the recent years a general discussion has started whether unsolved safety issues are related to advanced fuel. Advanced fuel is in this context a summary of features like high burnup, improved clad materials, low leakage loading pattern with high peaking factors etc. The design basis accidents RIA and Loca are of special interest for this discussion. From the Siemens point of view RIA is not a safety issue. There are sufficient margins between the enthalpy rise calculated by modern 3D methods and the fuel failures which occurred in RIA simulation tests when the effect of pulse width is taken into account. The evaluation of possible uncertainties for the established Loca criteria (17% equivalent corrosion, 1200 C clad temperature) for high burnup makes sense. But fuel with high burnup has significantly lower peaking factors than fuel with lower burnup. This gives sufficient margin counterbalancing possible uncertainties. In contrast to the above incomplete control rod insertion at higher burnup is potentially a real safety issue. Although Siemens fuel was not affected by the reported incidents they addressed the problem and checked that they have sufficient design margin for their fuel. (orig.) [de

Nuclear power plant control and instrumentation in Italy

International Nuclear Information System (INIS)

Lantieri, A.

1992-01-01

The National Energy Plan (PEN) approved by the Government of Italy in August 1988 provides a programme of research and industrial development or reactors with inherent and passive safety features. For the Control Systems and Instrumentation there is the aim to define rules and design criteria, by evaluating the impact of inherent safety goals on the C and I design. The effort on man-machine interface is considered essential to increase safety and efficiency of advanced reactors. The paper briefly describes the activity in control systems and in the instrumentation area. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-06-30

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

Specification of advanced safety modeling requirements (Rev. 0)

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Human factors aspects of advanced instrumentation in the nuclear industry

International Nuclear Information System (INIS)

Carter, R.J.

1989-01-01

An important consideration in regards to the use of advanced instrumentation in the nuclear industry is the interface between the instrumentation system and the human. A survey, oriented towards identifying the human factors aspects of digital instrumentation, was conducted at a number of United States (US) and Canadian nuclear vendors and utilities. Human factors issues, subsumed under the categories of computer-generated displays, controls, organizational support, training, and related topics were identified. 20 refs., 2 tabs

Advanced Instrumentation for Ultrafast Science at the LCLS

Energy Technology Data Exchange (ETDEWEB)

Berrah, Nora [Univ. of Connecticut, Storrs, CT (United States)

2015-10-13

This grant supported a Single Investigator and Small Group Research (SISGR) application to enable multi-user research in Ultrafast Science using the Linac Coherent Light Source (LCLS), the world’s first hard x-ray free electron laser (FEL) which lased for the first time at 1.5 Å on April 20, 2009. The goal of our proposal was to enable a New Era of Science by requesting funds to purchase and build Advanced Instrumentation for Ultrafast Science (AIUS), to utilize the intense, short x-ray pulses produced by the LCLS. The proposed instrumentation will allow peer review selected users to probe the ultrasmall and capture the ultrafast. These tools will expand on the investment already made in the construction of the light source and its instrumentation in both the LCLS and LUSI projects. The AIUS will provide researchers in the AMO, Chemical, Biological and Condensed Matter communities with greater flexibility in defining their scientific agenda at the LCLS. The proposed instrumentation will complement and significantly augment the present AMO instrument (funded through the LCLS project) through detectors and capabilities not included in the initial suite of instrumentation at the facility. We have built all of the instrumentations and they have been utilized by scientists. Please see report attached.

OASIS: An automotive analysis and safety engineering instrument

International Nuclear Information System (INIS)

Mader, Roland; Armengaud, Eric; Grießnig, Gerhard; Kreiner, Christian; Steger, Christian; Weiß, Reinhold

2013-01-01

In this paper, we describe a novel software tool named OASIS (AutOmotive Analysis and Safety EngIneering InStrument). OASIS supports automotive safety engineering with features allowing the creation of consistent and complete work products and to simplify and automate workflow steps from early analysis through system development to software development. More precisely, it provides support for (a) model creation and reuse, (b) analysis and documentation and (c) configuration and code generation. We present OASIS as a part of a tool chain supporting the application of a safety engineering workflow aligned with the automotive safety standard ISO 26262. In particular, we focus on OASIS' (1) support for property checking and model correction as well as its (2) support for fault tree generation and FMEA (Failure Modes and Effects Analysis) table generation. Finally, based on the case study of hybrid electric vehicle development, we demonstrate that (1) and (2) are able to strongly support FTA (Fault Tree Analysis) and FMEA

Advanced In-Pile Instrumentation for Materials Testing Reactors

Science.gov (United States)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

2014-08-01

The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

On the safety performance of the advanced nuclear energy systems

International Nuclear Information System (INIS)

Li Shounan

1999-01-01

Some features on the safety performances of the Advanced Nuclear Energy Systems are discussed. The advantages and some peculiar problems on the safety of Advanced Nuclear Energy Systems with subcritical nuclear reactor driven by external neutron sources are also pointed out in comparison with conventional nuclear reactors

Human factors survey of advanced instrumentation and controls

International Nuclear Information System (INIS)

Carter, R.J.

1989-01-01

A survey oriented towards identifying the human factors issues in regard to the use of advanced instrumentation and controls (I ampersand C) in the nuclear industry was conducted. A number of United States (US) and Canadian nuclear vendors and utilities were participants in the survey. Human factors items, subsumed under the categories of computer-generated displays (CGD), controls, organizational support, training, and related topics, were discussed. The survey found the industry to be concerned about the human factors issues related to the implementation of advanced I ampersand C. Fifteen potential human factors problems were identified. They include: the need for an advanced I ampersand C guideline equivalent to NUREG-0700; a role change in the control room from operator to supervisor; information overload; adequacy of existing training technology for advanced I ampersand C; and operator acceptance and trust. 11 refs., 1 tab

Passive safety and the advanced liquid metal reactors

International Nuclear Information System (INIS)

Hill, D.J.; Pedersen, D.R.; Marchaterre, J.F.

1988-01-01

Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental programs in this area

Instrumental variable methods in comparative safety and effectiveness research.

Science.gov (United States)

Brookhart, M Alan; Rassen, Jeremy A; Schneeweiss, Sebastian

2010-06-01

Instrumental variable (IV) methods have been proposed as a potential approach to the common problem of uncontrolled confounding in comparative studies of medical interventions, but IV methods are unfamiliar to many researchers. The goal of this article is to provide a non-technical, practical introduction to IV methods for comparative safety and effectiveness research. We outline the principles and basic assumptions necessary for valid IV estimation, discuss how to interpret the results of an IV study, provide a review of instruments that have been used in comparative effectiveness research, and suggest some minimal reporting standards for an IV analysis. Finally, we offer our perspective of the role of IV estimation vis-à-vis more traditional approaches based on statistical modeling of the exposure or outcome. We anticipate that IV methods will be often underpowered for drug safety studies of very rare outcomes, but may be potentially useful in studies of intended effects where uncontrolled confounding may be substantial.

Instrumental variable methods in comparative safety and effectiveness research†

Science.gov (United States)

Brookhart, M. Alan; Rassen, Jeremy A.; Schneeweiss, Sebastian

2010-01-01

Summary Instrumental variable (IV) methods have been proposed as a potential approach to the common problem of uncontrolled confounding in comparative studies of medical interventions, but IV methods are unfamiliar to many researchers. The goal of this article is to provide a non-technical, practical introduction to IV methods for comparative safety and effectiveness research. We outline the principles and basic assumptions necessary for valid IV estimation, discuss how to interpret the results of an IV study, provide a review of instruments that have been used in comparative effectiveness research, and suggest some minimal reporting standards for an IV analysis. Finally, we offer our perspective of the role of IV estimation vis-à-vis more traditional approaches based on statistical modeling of the exposure or outcome. We anticipate that IV methods will be often underpowered for drug safety studies of very rare outcomes, but may be potentially useful in studies of intended effects where uncontrolled confounding may be substantial. PMID:20354968

On safety classification of instrumentation and control systems and their components

International Nuclear Information System (INIS)

Yastrebenetskij, M.A.; Rozen, Yu.V.

2004-01-01

Safety classification of instrumentation and control systems (I and C) and their components (hardware, software, software-hardware complexes) is described: - evaluation of classification principles and criteria in Ukrainian standards and rules; comparison between Ukrainian and international principles and criteria; possibility and ways of coordination of Ukrainian and international standards related to (I and C) safety classification

Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description

International Nuclear Information System (INIS)

WHITEHURST, R.

1999-01-01

This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process

Regulatory instrument review: Management of aging of LWR [light water reactor] major safety-related components

International Nuclear Information System (INIS)

Werry, E.V.

1990-10-01

This report comprises Volume 1 of a review of US nuclear plant regulatory instruments to determine the amount and kind of information they contain on managing the aging of safety-related components in US nuclear power plants. The review was conducted for the US Nuclear Regulatory Commission (NRC) by the Pacific Northwest Laboratory (PNL) under the NRC Nuclear Plant Aging Research (NPAR) Program. Eight selected regulatory instruments, e.g., NRC Regulatory Guides and the Code of Federal Regulations, were reviewed for safety-related information on five selected components: reactor pressure vessels, steam generators, primary piping, pressurizers, and emergency diesel generators. Volume 2 will be concluded in FY 1991 and will also cover selected major safety-related components, e.g., pumps, valves and cables. The focus of the review was on 26 NPAR-defined safety-related aging issues, including examination, inspection, and maintenance and repair; excessive/harsh testing; and irradiation embrittlement. The major conclusion of the review is that safety-related regulatory instruments do provide implicit guidance for aging management, but include little explicit guidance. The major recommendation is that the instruments be revised or augmented to explicitly address the management of aging

Validity of instruments to assess students' travel and pedestrian safety.

Science.gov (United States)

Mendoza, Jason A; Watson, Kathy; Baranowski, Tom; Nicklas, Theresa A; Uscanga, Doris K; Hanfling, Marcus J

2010-05-18

Safe Routes to School (SRTS) programs are designed to make walking and bicycling to school safe and accessible for children. Despite their growing popularity, few validated measures exist for assessing important outcomes such as type of student transport or pedestrian safety behaviors. This research validated the SRTS school travel survey and a pedestrian safety behavior checklist. Fourth grade students completed a brief written survey on how they got to school that day with set responses. Test-retest reliability was obtained 3-4 hours apart. Convergent validity of the SRTS travel survey was assessed by comparison to parents' report. For the measure of pedestrian safety behavior, 10 research assistants observed 29 students at a school intersection for completion of 8 selected pedestrian safety behaviors. Reliability was determined in two ways: correlations between the research assistants' ratings to that of the Principal Investigator (PI) and intraclass correlations (ICC) across research assistant ratings. The SRTS travel survey had high test-retest reliability (kappa = 0.97, n = 96, p < 0.001) and convergent validity (kappa = 0.87, n = 81, p < 0.001). The pedestrian safety behavior checklist had moderate reliability across research assistants' ratings (ICC = 0.48) and moderate correlation with the PI (r = 0.55, p = < 0.01). When two raters simultaneously used the instrument, the ICC increased to 0.65. Overall percent agreement (91%), sensitivity (85%) and specificity (83%) were acceptable. These validated instruments can be used to assess SRTS programs. The pedestrian safety behavior checklist may benefit from further formative work.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Ualikhan Zhiyenbayev [KAIST, Daejeon (Korea, Republic of); Chung, Dae Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

International Nuclear Information System (INIS)

Ualikhan Zhiyenbayev; Chung, Dae Wook

2015-01-01

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether

Workshop on Advanced Technologies for Planetary Instruments, part 1

International Nuclear Information System (INIS)

Appleby, J.F.

1993-01-01

This meeting was conceived in response to new challenges facing NASA's robotic solar system exploration program. This volume contains papers presented at the Workshop on Advanced Technologies for Planetary Instruments on 28-30 Apr. 1993. This meeting was conceived in response to new challenges facing NASA's robotic solar system exploration program. Over the past several years, SDIO has sponsored a significant technology development program aimed, in part, at the production of instruments with these characteristics. This workshop provided an opportunity for specialists from the planetary science and DOD communities to establish contacts, to explore common technical ground in an open forum, and more specifically, to discuss the applicability of SDIO's technology base to planetary science instruments

Advanced Messaging Concept Development Basic Safety Message

Data.gov (United States)

Department of Transportation — Contains all Basic Safety Messages (BSMs) collected during the Advanced Messaging Concept Development (AMCD) field testing program. For this project, all of the Part...

Impact of proof test interval and coverage on probability of failure of safety instrumented function

International Nuclear Information System (INIS)

Jin, Jianghong; Pang, Lei; Hu, Bin; Wang, Xiaodong

2016-01-01

Highlights: • Introduction of proof test coverage makes the calculation of the probability of failure for SIF more accurate. • The probability of failure undetected by proof test is independently defined as P TIF and calculated. • P TIF is quantified using reliability block diagram and simple formula of PFD avg . • Improving proof test coverage and adopting reasonable test period can reduce the probability of failure for SIF. - Abstract: Imperfection of proof test can result in the safety function failure of safety instrumented system (SIS) at any time in its life period. IEC61508 and other references ignored or only elementarily analyzed the imperfection of proof test. In order to further study the impact of the imperfection of proof test on the probability of failure for safety instrumented function (SIF), the necessity of proof test and influence of its imperfection on system performance was first analyzed theoretically. The probability of failure for safety instrumented function resulted from the imperfection of proof test was defined as probability of test independent failures (P TIF ), and P TIF was separately calculated by introducing proof test coverage and adopting reliability block diagram, with reference to the simplified calculation formula of average probability of failure on demand (PFD avg ). Research results show that: the shorter proof test period and the higher proof test coverage indicate the smaller probability of failure for safety instrumented function. The probability of failure for safety instrumented function which is calculated by introducing proof test coverage will be more accurate.

Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

International Nuclear Information System (INIS)

Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I.; Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A.

2010-10-01

The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY TM platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY TM platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY TM platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1995-12-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1991-09-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Progress of nuclear safety for symbiosis and sustainability advanced digital instrumentation, control and information systems for nuclear power plants

CERN Document Server

Yoshikawa, Hidekazu

2014-01-01

This book introduces advanced methods of computational and information systems allowing readers to better understand the state-of-the-art design and implementation technology needed to maintain and enhance the safe operation of nuclear power plants. The subjects dealt with in the book are (i) Full digital instrumentation and control systems and human?machine interface technologies (ii) Risk? monitoring methods for large and? complex? plants (iii) Condition monitors for plant components (iv) Virtual and augmented reality for nuclear power plants and (v) Software reliability verification and val

1st Advanced Marine Renewable Energy Instrumentation Experts Workshop

Energy Technology Data Exchange (ETDEWEB)

none,

2011-10-01

The U.S. marine energy industry is actively pursuing development of offshore wind and marine hydrokinetic (MHK) energy systems. Experience in the wind energy sector demonstrates that new technology development requires thorough measurement and characterization of the environmental conditions prevalent at installation sites and of technology operating in the field. Presently, there are no turn-key instrumentation system solutions that meet the measurement needs of the marine energy industry. The 1st Advanced Marine Renewable Energy Instrumentation Experts Workshop brought together technical experts from government laboratories, academia, and industry representatives from marine energy, wind, offshore oil and gas, and instrumentation developers to present and discuss the instrumentation needs of the marine energy industry. The goals of the meeting were to: 1. Share the latest relevant knowledge among technical experts; 2. Review relevant state-of-the-art field measurement technologies and methods; 3. Review lessons learned from recent field deployments; 4. Identify synergies across different industries; 5. Identify gaps between existing and needed instrumentation capabilities; 6. Understand who are the leading experts; 7. Provide a forum where stakeholders from the marine energy industry could provide substantive input in the development of new marine energy field deployable instrumentation packages.

ATLAS program for advanced thermal-hydraulic safety research

International Nuclear Information System (INIS)

Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

2015-01-01

Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

Development of digital safety system logic and control

International Nuclear Information System (INIS)

Nishikawa, H.; Sakamoto, H.

1995-01-01

Advanced-BWR (ABWR) uses total digital control and instrumentation (C and I) system. In particular, ABWR adopts a newly developed safety system using advanced digital technology. In the presentation the digital safety system design, manufacturing and factory validation test method are shortly overviewed. The digital safety system consists of micro-processor based digital controllers, data and information transmission by optical fibers and human-machine interface using color flat displays. This new developed safety system meet the nuclear safety requirements such as high reliability, independence of divisions, operability and maintainability. (2 refs., 4 figs., 1 tab.)

Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

Energy Technology Data Exchange (ETDEWEB)

Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I. [Research and Production Corporation Radiy, 29 Geroev Stalingrada Str., Kirovograd 25006 (Ukraine); Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A., E-mail: marketing@radiy.co [Center for Safety Infrastructure-Oriented Research and Analysis, 37 Astronomicheskaya Str., Kharkiv 61085 (Ukraine)

2010-10-15

The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY{sup TM} platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY{sup TM} platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY{sup TM} platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)

Advanced In-pile Instrumentation for Material and Test Reactors

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

2013-06-01

The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

Physics and safety of advanced research reactors

International Nuclear Information System (INIS)

Boening, K.; Hardt, P. von der

1987-01-01

Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

Research on conceptual design of simplified nuclear safety instrument and control system

International Nuclear Information System (INIS)

Huang Jie

2015-01-01

The Nuclear safety instrument and control system is directly related to the safety of the reactor. So redundant and diversity design is used to ensure the system's security and reliability. This make the traditional safety system large, more cabinets and wiring complexity. To solve these problem, we can adopt new technology to make the design more simple. The simplify conceptual design can make the system less cabinets, less wiring, but high security, strong reliability. (author)

Westinghouse Advances in Passive Plant Safety

International Nuclear Information System (INIS)

Bruschi, H. J.; Manager, General; Gerstenhaber, E.

1993-01-01

On June 26, 1992, Westinghouse submitted the Ap600 Standard Safety Analysis Report and comprehensive PIRA results to the U. S. NRC for review as part of the Ap600 design certification program. This major milestone was met on time on a schedule set more than 3 years before submittal and is the result of the cooperative efforts of the U. S. Department of Energy (DOE), the Electric Power Requirements Program, and the Westinghouse Ap600 design team. These efforts were initiated in 1985 to develop a 600 MW advanced light water reactor plant design based on specific technical requirements established to provide the safety, simplicity, reliability, and economics necessary for the next generation of nuclear power plants. The Ap600 design achieves the ALRR safety requirements through ample design margins, simplified safety systems based on natural driving forces, and on a human-engineered man-machine interface system. Extensive Probabilistic Risk evolution, have recently shown that even if none of the active defense-in-depth safety systems are available, the passive systems alone meet safety goals. Furthermore, many tests in an extensive test program have begun or have been completed. Early tests show that passive safety perform well and meet design expectations

Application of life-cycle information for advancement in safety of nuclear fuel cycle facilities. Application of safety information to advanced safety management support system

International Nuclear Information System (INIS)

Suzuki, Kazuhiko; Ishida, Michihiko

2005-08-01

Risk management is major concern to nuclear energy reprocessing plants to improve plant and process reliability and ensure their safety. This is because we are required to predict potential risks before any accident or disaster occurs. The advancement of safety design and safety systems technologies showed large amount of useful safety-related knowledge that can be of great importance to plant operation to reduce operation risks and ensure safety. This research proposes safety knowledge modeling framework on the basis of ontology technologies to systematically construct plant knowledge model, which includes plant structure, operation, and the associated behaviors. In such plant knowledge model safety related information is defined and linked to the different elements of plant knowledge model. Ontology editor is employed to define the basic concepts and their inter-relations, which are used to capture and construct plant safety knowledge. In order to provide detailed safety knowledgebase, HAZOP results are analyzed and structured so that safety-related knowledge are identified and structured within the plant knowledgebase. The target safety knowledgebase includes: failures, deviations, causes, consequences, and fault propagation as mapped to plant knowledge. The proposed ontology-based safety framework is applied on case study nuclear plant to structure failures, causes, consequences, and fault propagation, which are used to support plant operation. (author)

Holistic safety analysis for advanced nuclear power plants

International Nuclear Information System (INIS)

Alvarenga, M.A.B.; Guimaraes, A.C.F.

1992-01-01

This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

Energy Technology Data Exchange (ETDEWEB)

J. L. Rempe; D. L. Knudson; J. E. Daw

2011-03-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Daw, J.E.

2011-01-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program's strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

Advanced Sensors for Safety and Security

CERN Document Server

Khudaverdyan, Surik

2013-01-01

This book results from a NATO Advanced Research Workshop titled “Technological Innovations in CBRNE Sensing and Detection for Safety, Security, and Sustainability” held in Yerevan, Armenia in 2012. The objective was to discuss and exchange views as to how fusion of advanced technologies can lead to improved sensors/detectors in support of defense, security, and situational awareness. The chapters range from policy and implementation, advanced sensor platforms using stand-off (THz and optical) and point-contact methods for detection of chemical, nuclear, biological, nuclear and explosive agents and contaminants in water, to synthesis methods for several materials used for sensors.  In view of asymmetric, kinetic, and distributed nature of threat vectors, an emphasis is placed to examine new generation of sensors/detectors that utilize an ecosystems of innovation and advanced sciences convergence in support of effective counter-measures against  CBRNE threats. The book will be of considerable interest and...

Recent technological advancements in laparoscopic surgical instruments

Science.gov (United States)

Subido, Edwin D. C.; Pacis, Danica Mitch M.; Bugtai, Nilo T.

2018-02-01

Laparoscopy was a progressive step to advancing surgical procedures as it minimised the scars left on the body after surgery, compared to traditional open surgery. Many years later, single-incision laparoscopic surgery (SILS) was created where, instead of having multiple incisions, only one incision is made or multiple small incisions in one location. SILS, or laparoendoscopic single-site surgery (LESS), may produce lesser scars but drawbacks for the surgeons are still present. This paper aims to present related literature of the recent technological developments in laparoscopic tools and procedure particularly in the vision system, handheld instruments. Tech advances in LESS will also be shown. Furthermore, this review intends to give an update on what has been going on in the surgical robot market and state which companies are interested and are developing robotic systems for commercial use to challenge Intuitive Surgical's da Vinci Surgical System that currently dominates the market.

Application of Safety Instrumented System (SIS) approach in older nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Nasimi, Elnara; Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca

2016-05-15

Highlights: • Study Safety Instrumented System (SIS) design for older nuclear power plant. • Apply SIS on Reheater Drains (RD) system. • Apply IEC 61508/61511 to design safety system. • Evaluate risk reduction based on proposed SIS design. - Abstract: In order to remain economically effective and financially profitable, the modern industries have to take their safety culture to a higher level and consider production losses in addition to simple accident prevention techniques. Ideally, compliance with safety requirements start during early design stages, but in some older facilities provisions for Safety Instrumented Systems (SIS) may not have been originally included. In this paper, a case study of a Reheater Drains (RD) system is used to illustrate such an example. Frequent failures of tank level controller lead to transients where the operation of shutting down RD pumps requires operators to manually isolate the quenching water and to close the main steam admission valves. Water in this system is at saturation temperature for the reheater steam side pressure, and any manual operation of the system is highly undesirable due to hazards of working with wet steam at approximately 758 kPa(g) pressure, preheated to 237 °C. Additionally, losses of inventory are highly undesirable as well and challenge other systems in the plant. In this paper, it is suggested that RD system can benefit from installation of an independent SIS system in order to address current challenges. This idea is being explored using IEC 61508 framework for “Functional safety of electrical/electronic/programmable electronic safety-related systems” to provide assurance that the SIS will offer the necessary risk reduction required to achieve required safety for the equipment.

Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

International Nuclear Information System (INIS)

2014-09-01

Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

Seismic isolation of Advanced LIGO: Review of strategy, instrumentation and performance

International Nuclear Information System (INIS)

Matichard, F; Mittleman, R; Mason, K; Biscans, S; Barnum, S; Evans, M; Foley, S; Lantz, B; Celerier, C; Clark, D; DeBra, D; Kissel, J; Allwine, E; Abbott, B; Abbott, R; Abbott, S; Coyne, D; McIver, J; Birch, J; DeRosa, R

2015-01-01

The new generation of gravitational waves detectors require unprecedented levels of isolation from seismic noise. This article reviews the seismic isolation strategy and instrumentation developed for the Advanced LIGO observatories. It summarizes over a decade of research on active inertial isolation and shows the performance recently achieved at the Advanced LIGO observatories. The paper emphasizes the scientific and technical challenges of this endeavor and how they have been addressed. An overview of the isolation strategy is given. It combines multiple layers of passive and active inertial isolation to provide suitable rejection of seismic noise at all frequencies. A detailed presentation of the three active platforms that have been developed is given. They are the hydraulic pre-isolator, the single-stage internal isolator and the two-stage internal isolator. The architecture, instrumentation, control scheme and isolation results are presented for each of the three systems. Results show that the seismic isolation sub-system meets Advanced LIGO’s stringent requirements and robustly supports the operation of the two detectors. (paper)

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...

Challenges in safety regulation of R and D activities for advanced technologies in DAE units

International Nuclear Information System (INIS)

Shukla, Dinesh Kumar

2016-01-01

DAE is engaged in intensive research and developmental activities, especially for advanced technologies such as accelerators, lasers, supercomputers, advanced materials and instrumentation. The starting point of an R and D project might be a hypothesis to be tested, problem to be solved, or the performance of an item to be improved, and there may be many possible solutions and technologies that could be used. R and D is quite different from designing, constructing, operating a plant. In these, precisely described result can be defined from the beginning and can be described in design specifications, process descriptions and procedures. However, while established procedures may be available to begin an R and D project, deviation from these procedures may occur often as a legitimate component of the conduct of R and D. Nevertheless, the R and D activities have to be performed in a manner which provides assurance that safety requirements are adequately addressed. Hence, the regulatory approach for enforcing the safety regulation in such facilities is also not as rigid as those for an operating industry. This paper is aimed to discuss some of the key challenges in regulating such R and D activities and also attempts to suggest a way forward. (author)

Experimental research progress on passive safety systems of Chinese advanced PWR

International Nuclear Information System (INIS)

Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan

2003-01-01

TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system

System 80+ instrumentation and controls - certification of a reliable design

International Nuclear Information System (INIS)

Matzie, R.A.; Scarola, K.; Turk, R.S.

1993-01-01

ABB Combustion Engineering's (ABB) System 80+ advanced light water plant design includes a modern, fully digitized instrumentation and controls complex, Nuplex 80+. This complex incorporates an evolutionary advanced control room, replacing conventional analog instruments with more capable computer driven components. As a result, Nuplex 80+ results in significant improvements in operator information handling and control to enhance plant safety and availability. The design implements features which the U.S. NRC has determined to be acceptable for addressing the potential for common mode failure in software implemented for protective functions. (author)

Instrumentation and control systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design

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

International Nuclear Information System (INIS)

Liu Sujuan

2002-01-01

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

Proceedings of the symposium on advanced measurement techniques and instrumentation: abstract book

International Nuclear Information System (INIS)

Kale, Y.B.; Kushwaha, M.; Somkuwar, S.P.; Ajayakumar, S.; Sampathkumar, R.

2011-01-01

In order to consolidate the existing knowledge base and further to focus on the future directions of the field of advanced measurement techniques and instrumentation, Bhabha Atomic Research Centre has organized a three-day symposium on 'Advanced Measurement Techniques and Instrumentation' at Multi Purpose Hall, Training School Hostel, Anushaktinagar, Mumbai during February 02-04, 2011. The symposium is aimed at providing a forum to discuss the emerging trends and challenges ahead in the important area of measurement science and technology. This is a unique symposium, which brings together scientists and engineers from all disciplines and provides them a platform for close interaction to exchange ideas, methodologies and expertise, which is extremely important for synergic growth of this field. The symposium consists of 27 talks, which include keynote address, plenary and invited talks, and 63 contributory papers. The abstracts of these papers are brought to you in this volume. Readers may observe that the scientific programme of the symposium covers a wide ranging issues including advanced scientific concepts in measurements, instrumentation strategies, mathematical techniques and development of devices for applications in fundamental physics, astrophysics, fusion plasmas, nuclear reactors, accelerators, environment, chemical and biological sciences, and national security. Papers relevant to INIS are indexed separately

Simultaneous control of multiple instruments at the Advanced Technology Solar Telescope

Science.gov (United States)

Johansson, Erik M.; Goodrich, Bret

2012-09-01

The Advanced Technology Solar Telescope (ATST) is a 4-meter solar observatory under construction at Haleakala, Hawaii. The simultaneous use of multiple instruments is one of the unique capabilities that makes the ATST a premier ground based solar observatory. Control of the instrument suite is accomplished by the Instrument Control System (ICS), a layer of software between the Observatory Control System (OCS) and the instruments. The ICS presents a single narrow interface to the OCS and provides a standard interface for the instruments to be controlled. It is built upon the ATST Common Services Framework (CSF), an infrastructure for the implementation of a distributed control system. The ICS responds to OCS commands and events, coordinating and distributing them to the various instruments while monitoring their progress and reporting the status back to the OCS. The ICS requires no specific knowledge about the instruments. All information about the instruments used in an experiment is passed by the OCS to the ICS, which extracts and forwards the parameters to the appropriate instrument controllers. The instruments participating in an experiment define the active instrument set. A subset of those instruments must complete their observing activities in order for the experiment to be considered complete and are referred to as the must-complete instrument set. In addition, instruments may participate in eavesdrop mode, outside of the control of the ICS. All instrument controllers use the same standard narrow interface, which allows new instruments to be added without having to modify the interface or any existing instrument controllers.

Advance on solar instrumentation in China

Science.gov (United States)

Yan, Yihua

2015-08-01

The solar observing facilities in China are introduced with the emphasis on the development in recent years and future plans for both ground and space-based solar instrumentations. The recent solar instruments are as follows: A new generation Chinese Spectral Radioreliograph (CSRH) has been constructed at Mingantu Observing Station in Zhengxiangbaiqi, inner Mongolia of China since 2013 and is in test observations now. CSRH has two arrays with 40 × 4.5 m and 60 × 2 m parabolic antennas covering 0.4-2 GHz and 2-15 GHz frequency range. CSRH is renamed as MUSER (Mingantu Ultrawide Spectral Radiheliograph) after its accomplishment. A new 1 m vacuum solar telescope (NVST) has been installed in 2010 at Fuxian lake, 60 km away from Kunming, Yunana. At present it is the best seeing place in China. A new telescope called ONSET (Optical and NIR Solar Eruption Tracer) has been established at the same site as NVST in 2011. ONSET has been put into operation since 2013. For future ground-based plans, Chinese Giant Solar Telescope (CGST) with spatial resolution equivalent to 8m and effective area of 5m full-aperture telescope has been proposed and was formally listed into the National Plans of Major Science & Technology Infrastructures in China. The pre-study and site survey for CGST have been pursued. A 1-meter mid-infrared telescope for precise measurement of the solar magnetic field has been funded by NSFC in 2014 as a national major scientific instrument development project. This project will develop the first mid-infrared solar magnetic observation instrument in the world aiming at increasing the precision of the transverse magnetic field measurement by one order of magnitude. For future ground-based plans, we promote the Deep-space Solar Observatory (DSO) with 1-m aperture telescope to be formally funded. The ASO-S (an Advanced Space-based Solar Observatory) has been supported in background phase by Space Science Program as a small mission. Other related space solar

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1989-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

Handbook of advanced nuclear hydrogen safety. 1st edition

International Nuclear Information System (INIS)

Hino, Ryutaro; Takegami, Hiroaki; Ogawa, Toru

2017-03-01

In the aftermath of the Fukushima nuclear accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. Therefore, efforts have begun to form a common knowledge base between nuclear engineers and experts on combustion and explosion, and to secure and improve future nuclear energy safety. As one of such activities, we have prepared the 'Handbook of Advanced Nuclear Hydrogen Safety'. A handbook committee consisting of Japanese experts in the fields of nuclear and combustion-explosion in universities, nuclear companies, electric companies and research institutes was established in 2012. The objective and consents of the handbook were determined, and the outline of the contents was decided. The concepts of the handbook are as follows: to show advanced nuclear hydrogen safety technologies that nuclear engineers should understand, to show hydrogen safety points to make combustion-explosion experts cooperate with nuclear engineers, to expand information on water radiolysis considering the situation from just after the Fukushima accidents and to the waste management necessary for decommissioning after the accident etc. Many experts have participated to manuscript preparation, which was the first step of forming a hydrogen community across the boundaries of fields. The hydrogen community is expected to grow along with its improvement to the knowledge base on nuclear hydrogen safety. (author)

Considerations on monitoring needs of advanced, passive safety light water reactors for severe accident management

International Nuclear Information System (INIS)

Bava, G.; Zambardi, F.

1992-01-01

This paper deals with problems concerning information and related instrumentation needs for Accident Management (AM), with special emphasis on Severe Accidents (SA) in the new advanced, passive safety Light Water Reactors (PLWR), presently in a development stage. The passive safety conception adopted in the plants concerned goes parallel with a deeper consideration of SA, that reflects the need of increasing the plant resistance against conditions going beyond traditional ''design basis accidents''. Further, the role of Accident Management (AM) is still emphasized as last step of the defence in depth concept, in spite of the design efforts aimed to reduce human factor importance; as a consequence, the availability of pertinent information on actual plant conditions remains a necessary premise for performing preplanned actions. This information is essential to assess the evolution of the accident scenarios, to monitor the performances of the safety systems, to evaluate the ultimate challenge to the plant safety, and to implement the emergency operating procedures and the emergency plans. Based on these general purposes, the impact of the new conception on the monitoring structure is discussed, furthermore reference is made to the accident monitoring criteria applied in current plants to evaluate the requirements for possible solutions. (orig.)

Incorporation of advanced accident analysis methodology into safety analysis reports

International Nuclear Information System (INIS)

2003-05-01

The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

Advances in Predictive Toxicology for Discovery Safety through High Content Screening.

Science.gov (United States)

Persson, Mikael; Hornberg, Jorrit J

2016-12-19

High content screening enables parallel acquisition of multiple molecular and cellular readouts. In particular the predictive toxicology field has progressed from the advances in high content screening, as more refined end points that report on cellular health can be studied in combination, at the single cell level, and in relatively high throughput. Here, we discuss how high content screening has become an essential tool for Discovery Safety, the discipline that integrates safety and toxicology in the drug discovery process to identify and mitigate safety concerns with the aim to design drug candidates with a superior safety profile. In addition to customized mechanistic assays to evaluate target safety, routine screening assays can be applied to identify risk factors for frequently occurring organ toxicities. We discuss the current state of high content screening assays for hepatotoxicity, cardiotoxicity, neurotoxicity, nephrotoxicity, and genotoxicity, including recent developments and current advances.

Design of Instrumentation and Control Systems for Nuclear Power Plants. Specific Safety Guide

International Nuclear Information System (INIS)

2016-01-01

This publication is a revision and combination of two Safety Guides, IAEA Safety Standards Series No. NS-G-1.1 and No. NS-G-1.3. The revision takes into account developments in instrumentation and control (I&C) systems since the publication of the earlier Safety Guides. The main changes relate to the continuing development of computer applications and the evolution of the methods necessary for their safe, secure and practical use. In addition, account is taken of developments in human factors engineering and the need for computer security. This Safety Guide references and takes into account other IAEA Safety Standards and Nuclear Security Series publications that provide guidance relating to I&C design

Validity of instruments to assess students' travel and pedestrian safety

Directory of Open Access Journals (Sweden)

Baranowski Tom

2010-05-01

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

Halden Reactor Project Workshop: Understanding Advanced Instrumentation and Controls Issues

International Nuclear Information System (INIS)

Beltracchi, L.

1991-01-01

A Halden Reactor Project Workshop on 'Understanding Advanced Instrumentation and Controls Issues' was held in Halden, Norway, during June 17-18, 1991. The objectives of the workshop were to (1) identify and prioritize the types of technical information that the Halden Project can produce to facilitate the development of man-machine interface guidelines and (2) to identify methods to effectively integrate and disseminate this information to signatory organizations. As a member of the Halden Reactor Project, the Nuclear Regulatory Commission (NRC) requested the workshop. This request resulted from the NRC's need for human factors guidelines for the evaluation of advanced instrumentation and controls. The Halden Reactor Project is a cooperative agreement among several countries belonging to the Organization for Economic Cooperation and Development (OECD). The US began its association with the Halden Project in 1958 through the Atomic Energy Commission. The project's activities are centered at the Halden heavy-water reactor and its associated man-machine laboratory in Halden, Norway. The research program conducted at Halden consists of studies on fuel performance and computer-based man-machine interfaces

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

2014-02-01

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

Performance and safety design of the advanced liquid metal reactor

International Nuclear Information System (INIS)

Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

1991-01-01

The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

The qualification of electrical components and instrumentations relevant to safety

CERN Document Server

Zambardi, F

1989-01-01

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and ...

The advanced neutron source facility: Safety philosophy and studies

International Nuclear Information System (INIS)

Greene, S.R.; Harrington, R.M.

1988-01-01

The Advanced Neutron Source (ANS) is currently the only new civilian nuclear reactor facility proposed for construction in the United States. Even though the thermal power of this research-oriented reactor is a relatively low 300 MW, the design will undoubtedly receive intense scrutiny before construction is allowed to proceed. Safety studies are already under way to ensure that the maximum degree of safety in incorporated into the design and that the design is acceptable to the Department of Energy (DOE) and can meet the Nuclear Regulatory Commission regulations. This document discusses these safety studies

Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description SYS 93-2

International Nuclear Information System (INIS)

WHITEHURST, R.

1999-01-01

This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process

Recent advances in systems safety and security

CERN Document Server

Stamatescu, Grigore

2016-01-01

This book represents a timely overview of advances in systems safety and security, based on selected, revised and extended contributions from the 2nd and 3rd editions of the International Workshop on Systems Safety and Security – IWSSS, held in 2014 and 2015, respectively, in Bucharest, Romania. It includes 14 chapters, co-authored by 34 researchers from 7 countries. The book provides an useful reference from both theoretical and applied perspectives in what concerns recent progress in this area of critical interest. Contributions, broadly grouped by core topic, address challenges related to information theoretic methods for assuring systems safety and security, cloud-based solutions, image processing approaches, distributed sensor networks and legal or risk analysis viewpoints. These are mostly accompanied by associated case studies providing additional practical value and underlying the broad relevance and impact of the field.

Proposed plan for the development of advanced instrumentation and control technology in Korea

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Ham, Chang-Shik

1995-01-01

All of the nuclear power plants in Korea are operating with analog instrumentation and control (I and C) equipment which are increasingly faced with frequent troubles, obsolescence and high maintenance expense. Electrical and computer technology has improved rapidly in recent years and has been applied to other industries. So it is strongly recommended to adopt the modern digital and computer technology to improve plant safety and availability. The plan, which is aimed at replacement of existing I and C systems, and at improving planned as well as next generation digital I and C systems is divided into three major parts: (1) Plan for domestic design of I and C systems and components, (2) Plan for domestic manufacturing of I and C equipment, and (3) Plan for development of future technologies. This plan provides advanced digital I and C requirements equivalent to Electric Power Research Institute Utility Requirements Document (Chapter 10) Man Machine Interface Systems. Also this plan includes the critical issue of digital I and C, namely software verification and validation strategy. (6 refs., 1 fig., 2 tabs.)

Safety Assessment of Two Hybrid Instrumentation Techniques in a Dental Student Endodontic Clinic: A Retrospective Study.

Science.gov (United States)

Coelho, Marcelo Santos; Card, Steven John; Tawil, Peter Zahi

2017-03-01

The aim of this study was to retrospectively assess the safety potential of a hybrid technique combining nickel-titanium (NiTi) reciprocating and rotary instruments by third- and fourth-year dental students in the predoctoral endodontics clinic at one U.S. dental school. For the study, 3,194 root canal treatments performed by 317 dental students from 2012 through 2015 were evaluated for incidence of ledge creation and instrument separation. The hybrid reciprocating and rotary technique (RRT) consisted of a glide path creation with stainless steel hand files up to size 15/02, a crown down preparation with a NiTi reciprocating instrument, and an apical preparation with NiTi rotary instruments. The control was a traditional rotary and hand technique (RHT) that consisted of the same glide path procedure followed by a crown down preparation with NiTi rotary instruments and an apical preparation with NiTi hand instruments. The results showed that the RHT technique presented a rate of ledge creation of 1.4% per root and the RRT technique was 0.5% per root (protary technique for root canal instrumentation by these dental students provided good safety. This hybrid technique offered a low rate of ledge creation along with no NiTi instrument separation.

Advancement on safety management system of nuclear power for safety and non-anxiety of society

International Nuclear Information System (INIS)

Yoshikawa, Hidekazu

2004-01-01

Advancement on safety management system is investigated to improve safety and non-anxiety of society for nuclear power, from the standpoint of human machine system research. First, the recent progress of R and D works of human machine interface technologies since 1980 s are reviewed and then the necessity of introducing a new approach to promote technical risk communication activity to foster safety culture in nuclear industries. Finally, a new concept of Offsite Operation and Maintenance Support Center (OMSC) is proposed as the core facility to assemble human resources and their expertise in all organizations of nuclear power, for enhancing safety and non-anxiety of society for nuclear power. (author)

Probabilistic safety assessment for instrumentation and control systems in nuclear power plants. A literature survey

International Nuclear Information System (INIS)

Lu, Lixuan; Jiang, Jin

2003-01-01

Deregulation in electricity market will create a great deal of challenges for Nuclear Power Plants (NPP). To stay competitive, NPP will need to find new ways to reduce their operation costs. In NPP, Instrumentation and Control (I and C) systems play an important role in reducing the cost of producing electricity while maintaining and/or enhancing safety. Therefore, it is extremely important that one should manage the I and C systems more efficiently and economically. Meanwhile, obsolescence problem associated with I and C systems encouraged the usage of advanced digital techniques in I and C systems. Thus, new methodologies are needed to analyze the reliability and determine the maintenance strategy for the digital I and C systems. Probabilistic Safety Assessment (PSA) has been probed to be a promising method to deal with this issue. This paper provides a literature survey on the development of digital I and C systems in NPP, followed by a detailed review of PSA including its benefits, limitations and the future direction of its development. Most importantly, potential applications of PSA in various aspects of I and C systems are brought into perspective throughout the paper. Furthermore, the applicability of PSA in the regulation of safety-related I and C systems is demonstrated. Detailed information on PSA applications in 1) the resource allocation for I and C systems: 2) the determination of surveillance testing strategies; and 3) I and C system designs, is provided. (author)

Pressure Safety: Advanced Self-Study 30120

Energy Technology Data Exchange (ETDEWEB)

Glass, George [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-02-29

Pressure Safety Advance Self-Study (Course 30120) consists of an introduction, five modules, and a quiz. To receive credit in UTrain for completing this course, you must score 80% or better on the 15-question quiz (check UTrain). Directions for initiating the quiz are appended to the end of this training manual. This course contains several links to LANL websites. UTrain might not support active links, so please copy links into the address line in your browser.

The Advanced Light Source (ALS) Radiation Safety System

International Nuclear Information System (INIS)

Ritchie, A.; Oldfather, D.; Lindner, A.

1993-05-01

The Advanced Light Source (ALS) at the Lawrence Berkeley Laboratory (LBL) is a 1.5 GeV synchrotron light source facility consisting of a 120 keV electron gun, 50 MeV linear accelerator, 1.5 Gev booster synchrotron, 200 meter circumference electron storage ring, and many photon beamline transport systems for research. The Radiation Safety System for the ALS has been designed and built with a primary goal of providing protection against inadvertent personnel exposure to gamma and neutron radiation and, secondarily, to enhance the electrical safety of select magnet power supplies

Probabilistic safety assessment for instrumentation and control systems in nuclear power plants: an overview

International Nuclear Information System (INIS)

Lu, Lixuan; Jiang, Jin

2004-01-01

Deregulation in the electricity market has resulted in a number of challenges in the nuclear power industry. Nuclear power plants must find innovative ways to remain competitive by reducing operating costs without jeopardizing safety. Instrumentation and Control (I and C) systems not only play important roles in plant operation, but also in reducing the cost of power generation while maintaining and/or enhancing safety. Therefore, it is extremely important that I and C systems are managed efficiently and economically. With the increasing use of digital technologies, new methods are needed to solve problems associated with various aspects of digital I and C systems. Probabilistic Safety Assessment (PSA) has proved to be an effective method for safety analysis and risk-based decisions, even though challenges are still present. This paper provides an overview of PSA applications in three areas of digital I and C systems in nuclear power plants. These areas are Graded Quality Assurance, Surveillance Testing, and Instrumentation and Control System Design. In addition, PSA application in the regulation of nuclear power plants that adopt digital I and C systems is also investigated. (author)

Advanced research workshop: nuclear materials safety

International Nuclear Information System (INIS)

Jardine, L J; Moshkov, M M.

1999-01-01

The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

R and D perspectives on the advanced nuclear safety regulation system

International Nuclear Information System (INIS)

Lee, Chang Ju; Ahn, Sang Kyu; Park, Jong Seuk; Chung, Dae Wook; Han, Sang Hoon; Lee, Jung Won

2009-01-01

As current licensing process is much desired to be optimized both plant safety and regulatory efficiency, an advanced safety regulation such as risk informed regulation has been come out. Also, there is a need to have a future oriented safety regulation since a lot of new reactors are conceptualized. Keeping pace with these needs, since early 2007, Korean government has launched a new project for preparing an advanced and future oriented nuclear safety regulation system. In order to get practical achievements, the project team sets up such specific research objectives for the development of: implementation program for graded regulation using risk and performance information; multi purpose PSA models for regulatory uses; a technology neutral regulatory framework for future innovative reactors; evaluation procedure of proliferation resistance; and, performance based fire hazard analysis method and evaluation system. This paper introduces major R and D outputs of this project, and provides some perspectives for achieving effectiveness and efficiency of the nuclear regulation system in Korea

R and D perspectives on the advanced nuclear safety regulation system

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang Ju; Ahn, Sang Kyu; Park, Jong Seuk; Chung, Dae Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Han, Sang Hoon; Lee, Jung Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-04-15

As current licensing process is much desired to be optimized both plant safety and regulatory efficiency, an advanced safety regulation such as risk informed regulation has been come out. Also, there is a need to have a future oriented safety regulation since a lot of new reactors are conceptualized. Keeping pace with these needs, since early 2007, Korean government has launched a new project for preparing an advanced and future oriented nuclear safety regulation system. In order to get practical achievements, the project team sets up such specific research objectives for the development of: implementation program for graded regulation using risk and performance information; multi purpose PSA models for regulatory uses; a technology neutral regulatory framework for future innovative reactors; evaluation procedure of proliferation resistance; and, performance based fire hazard analysis method and evaluation system. This paper introduces major R and D outputs of this project, and provides some perspectives for achieving effectiveness and efficiency of the nuclear regulation system in Korea.

OECD/CSNI specialist meeting on advanced instrumentation and measurements techniques: summary and conclusions

International Nuclear Information System (INIS)

1997-01-01

This specialist meeting on Advanced Instrumentation and Measurements Techniques was held in Santa Barbara (USA) in 1997 and attracted some 70 participants in ten technical sessions and a session of the round table discussions, with a total of 41 papers. It was intended to bring together the international experts in multi-phase flow instrumentation, experiment and modeling to review the state-of-the-art of the two-phase flow instrumentation methods and to discuss the relation between modeling needs and instrumentation capabilities. The following topics were included: Modeling needs and future direction for improved constitutive relations, interfacial area transport equation, and multi-dimensional two-fluid model formulation; local instrumentation developments for void fraction, interfacial area, phase velocities, turbulence, entrainment, particle size, thermal non-equilibrium, shear stress, nucleation, condensation and boiling; global instrumentation developments for void fraction, mass flow, two-phase level, non-condensable concentration, flow regimes, low flow and break flow; relation between modeling needs and instrumentation capabilities, future directions for experiments focused on modeling needs and for instrumentation developments

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Advancements in Actuated Musical Instruments

DEFF Research Database (Denmark)

Overholt, Daniel; Berdahl, Edgar; Hamilton, Robert

2011-01-01

are physical instruments that have been endowed with virtual qualities controlled by a computer in real-time but which are nevertheless tangible. These instruments provide intuitive and engaging new forms of interaction. They are different from traditional (acoustic) and fully automated (robotic) instruments...

Development of an instrument to measure self-efficacy in caregivers of people with advanced cancer.

Science.gov (United States)

Ugalde, Anna; Krishnasamy, Meinir; Schofield, Penelope

2013-06-01

Informal caregivers of people with advanced cancer experience many negative impacts as a result of their role. There is a lack of suitable measures specifically designed to assess their experience. This study aimed to develop a new measure to assess self-efficacy in caregivers of people with advanced cancer. The development and testing of the new measure consisted of four separate, sequential phases: generation of issues, development of issues into items, pilot testing and field testing. In the generation of issues, 17 caregivers were interviewed to generate data. These data were analysed to generate codes, which were then systematically developed into items to construct the instrument. The instrument was pilot tested with 14 health professionals and five caregivers. It was then administered to a large sample for field testing to establish the psychometric properties, with established measures including the Brief Cope and the Family Appraisals for Caregiving Questionnaire for Palliative Care. Ninety-four caregivers completed the questionnaire booklet to establish the factor structure, reliability and validity. The factor analysis resulted in a 21-item, four-factor instrument, with the subscales being termed Resilience, Self-Maintenance, Emotional Connectivity and Instrumental Caregiving. The test-retest reliability and internal consistency were both excellent, ranging from 0.73 to 0.85 and 0.81 to 0.94, respectively. Six convergent and divergent hypotheses were made, and five were supported. This study has developed a new instrument to assess self-efficacy in caregivers of people with advanced cancer. The result is a four-factor, 21-item instrument with demonstrated reliability and validity. Copyright © 2012 John Wiley & Sons, Ltd.

Safety issues on advanced fuel

Energy Technology Data Exchange (ETDEWEB)

Gross, H.; Krebs, W.D. [Siemens AG, Bereich Energieerzeugug (KWU), Erlangen (Germany). Geschaeftsgebiet Nukleare Energieerzeugung

1998-05-01

In the recent years a general discussion has started whether unsolved safety issues are related to advanced fuel. Advanced fuel is in this context a summary of features like high burnup, improved clad materials, low leakage loading pattern with high peaking factors etc. The design basis accidents RIA and Loca are of special interest for this discussion. From the Siemens point of view RIA is not a safety issue. There are sufficient margins between the enthalpy rise calculated by modern 3D methods and the fuel failures which occurred in RIA simulation tests when the effect of pulse width is taken into account. The evaluation of possible uncertainties for the established Loca criteria (17% equivalent corrosion, 1200 C clad temperature) for high burnup makes sense. But fuel with high burnup has significantly lower peaking factors than fuel with lower burnup. This gives sufficient margin counterbalancing possible uncertainties. In contrast to the above incomplete control rod insertion at higher burnup is potentially a real safety issue. Although Siemens fuel was not affected by the reported incidents they addressed the problem and checked that they have sufficient design margin for their fuel. (orig.) [Deutsch] In den letzten Jahren hat eine allgemeine Diskussion begonnen, ob mit fortgeschrittenen Brennelementen (BE) ungeklaerte Sicherheitsprobleme verbunden sind. Dabei ist `Fortgeschrittene Brennelemente` ein Sammelbegriff fuer hohe Abbraende, verbesserte Huellrohrmaterialien, Low-leakage-Einsatzplanungen mit hohen Heissstellenfaktoren usw. Die Auslegungsstoerfaelle RIA und Loca sind in dieser Diskussion von besonderer Bedeutung. Aus der Sicht von Siemens ist der RIA kein Sicherheitsproblem. Zwischen den mit modernen 3D-Methoden berechneten Enthalpieerhoehungen und den in RIA-Experimenten aufgetretenen Brennstabdefekten bestehen ausreichende Abstaende, wenn der Einfluss der Pulsbreite beruecksichtigt wird. Die Untersuchung eventueller Unsicherheiten bei hohen

Safety and function of a new clinical intracerebral microinjection instrument for stem cells and therapeutics examined in the Göttingen minipig

DEFF Research Database (Denmark)

Bjarkam, Carsten R; GLUD, AN; Margolin, Lee

2010-01-01

Safety and function of a new clinical intracerebral microinjection instrument for stem cells and therapeutics examined in the Göttingen minipig......Safety and function of a new clinical intracerebral microinjection instrument for stem cells and therapeutics examined in the Göttingen minipig...

Instruments for non-destructive evaluation of advanced test reactor inpile tubes

International Nuclear Information System (INIS)

Livingston, R.A.; Beller, L.S.; Edgett, S.M.

1986-01-01

The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

Human-system safety methods for development of advanced air traffic management systems

International Nuclear Information System (INIS)

Nelson, William R.

1999-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) is supporting the National Aeronautics and Space Administration in the development of advanced air traffic management (ATM) systems as part of the Advanced Air Transportation Technologies program. As part of this program INEEL conducted a survey of human-system safety methods that have been applied to complex technical systems, to identify lessons learned from these applications and provide recommendations for the development of advanced ATM systems. The domains that were surveyed included offshore oil and gas, commercial nuclear power, commercial aviation, and military. The survey showed that widely different approaches are used in these industries, and that the methods used range from very high-level, qualitative approaches to very detailed quantitative methods such as human reliability analysis (HRA) and probabilistic safety assessment (PSA). In addition, the industries varied widely in how effectively they incorporate human-system safety assessment in the design, development, and testing of complex technical systems. In spite of the lack of uniformity in the approaches and methods used, it was found that methods are available that can be combined and adapted to support the development of advanced air traffic management systems (author) (ml)

Out-pile test of non-instrumented capsule for the advanced PWR fuel pellets in HANARO irradiation test

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Lee, C. B.; Oh, D. S.; Bang, J. K.; Kim, Y. M.; Yang, Y. S.; Jeong, Y. H.; Jeon, H. K.; Ryu, J. S. [KAERI, Taejon (Korea, Republic of)

2002-05-01

Non-instrumental capsule were designed and fabricated to irradiate the advanced pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. This capsule was out-pie tested at Cold Test Loop-I in KAERI. From the pressure drop test results, it is noted that the flow velocity across the non-instrumented capsule of advanced PWR fuel pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the capsule ranges from 13.0 to 32.3 Hz. RMS displacement for non-instrumented capsule of advanced PWR fuel pellet is less than 11.6 {mu}m, and the maximum displacement is less that 30.5 {mu}m. The flow rate for endurance test were 8.19 kg/s, which was 110% of 7.45 kg/s. And the endurance test was carried out for 100 days and 17 hours. The test results found not to the wear satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented capsule.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

T. A. Tomberlin; S. B. Grover

2004-11-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

T. A. Tomberlin; S. B. Grover

2004-01-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

Advanced nuclear reactor safety design technology research in NPIC

International Nuclear Information System (INIS)

Yu, H.

2014-01-01

After the Fukushima accident happen, Nuclear Power Plants (NPPs) construction has been suspended in China for a time. Now the new regulatory rule has been proposed that the most advanced safety standard must be adopted for the new NPPs and practical elimination of large fission product release by design during the next five plans period. So the advanced reactor research is developing in China. NPIC is engaging on the ACP1000 and ACP100 (Small Module Reactor) design. The main design character will be introduced in this paper. The Passive Combined with Active (PCWA) design was adopted during the ACP1000 design to reduce the core damage frequency (CDF); the Cavity Injection System (CIS) is design to mitigation the consequence of the severe accident. Advance passive safety system was designed to ensure the long term residual heat removal during the Small Module Reactor (SMR). The SMR will be utilized to be the floating reactors, district heating reactor and so on. Besides, the Science and Technology on Reactor System Design Technology Laboratory (LRSDT) also engaged on the fundamental thermal-hydraulic characteristic research in support of the system validation. (author)

Experience related to the safety of advanced LMFBR fuel elements

International Nuclear Information System (INIS)

Kerrisk, J.F.

1975-07-01

Experiments and experience relative to the safety of advanced fuel elements for the liquid metal fast breeder reactor are reviewed. The design and operating parameters and some of the unique features of advanced fuel elements are discussed breifly. Transient and steady state overpower operation and loss of sodium bond tests and experience are discussed in detail. Areas where information is lacking are also mentioned

PET/CT: underlying physics, instrumentation, and advances.

Science.gov (United States)

Torres Espallardo, I

Since it was first introduced, the main goal of PET/CT has been to provide both PET and CT images with high clinical quality and to present them to radiologists and specialists in nuclear medicine as a fused, perfectly aligned image. The use of fused PET and CT images quickly became routine in clinical practice, showing the great potential of these hybrid scanners. Thanks to this success, manufacturers have gone beyond considering CT as a mere attenuation corrector for PET, concentrating instead on design high performance PET and CT scanners with more interesting features. Since the first commercial PET/CT scanner became available in 2001, both the PET component and the CT component have improved immensely. In the case of PET, faster scintillation crystals with high stopping power such as LYSO crystals have enabled more sensitive devices to be built, making it possible to reduce the number of undesired coincidence events and to use time of flight (TOF) techniques. All these advances have improved lesion detection, especially in situations with very noisy backgrounds. Iterative reconstruction methods, together with the corrections carried out during the reconstruction and the use of the point-spread function, have improved image quality. In parallel, CT instrumentation has also improved significantly, and 64- and 128-row detectors have been incorporated into the most modern PET/CT scanners. This makes it possible to obtain high quality diagnostic anatomic images in a few seconds that both enable the correction of PET attenuation and provide information for diagnosis. Furthermore, nowadays nearly all PET/CT scanners have a system that modulates the dose of radiation that the patient is exposed to in the CT study in function of the region scanned. This article reviews the underlying physics of PET and CT imaging separately, describes the changes in the instrumentation and standard protocols in a combined PET/CT system, and finally points out the most important

Development of an instrument for the evaluation of advanced life support performance.

Science.gov (United States)

Peltonen, L-M; Peltonen, V; Salanterä, S; Tommila, M

2017-10-01

Assessing advanced life support (ALS) competence requires validated instruments. Existing instruments include aspects of technical skills (TS), non-technical skills (NTS) or both, but one instrument for detailed assessment that suits all resuscitation situations is lacking. This study aimed to develop an instrument for the evaluation of the overall ALS performance of the whole team. This instrument development study had four phases. First, we reviewed literature and resuscitation guidelines to explore items to include in the instrument. Thereafter, we interviewed resuscitation team professionals (n = 66), using the critical incident technique, to determine possible additional aspects associated with the performance of ALS. Second, we developed an instrument based on the findings. Third, we used an expert panel (n = 20) to assess the validity of the developed instrument. Finally, we revised the instrument based on the experts' comments and tested it with six experts who evaluated 22 video recorded resuscitations. The final version of the developed instrument had 69 items divided into adherence to guidelines (28 items), clinical decision-making (5 items), workload management (12 items), team behaviour (8 items), information management (6 items), patient integrity and consideration of laymen (4 items) and work routines (6 items). The Cronbach's α values were good, and strong correlations between the overall performance and the instrument were observed. The instrument may be useful for detailed assessment of the team's overall performance, but the numerous items make the use demanding. The instrument is still under development, and more research is needed to determine its psychometric properties. © 2017 The Acta Anaesthesiologica Scandinavica Foundation. Published by John Wiley & Sons Ltd.

The Advanced Light Source (ALS) Radiation Safety System

International Nuclear Information System (INIS)

Ritchie, A.L.; Oldfather, D.E.; Lindner, A.F.

1993-08-01

The Advanced Light Source (ALS) at the Lawrence Berkeley Laboratory (LBL) is a 1.5 Gev synchrotron light source facility consisting of a 120 kev electron gun, 50 Mev linear accelerator, 1.5 Gev booster synchrotron, 200 meter circumference electron storage ring, and many photon beamline transport systems for research. Figure 1. ALS floor plan. Pairs of neutron and gamma radiation monitors are shown as dots numbered from 1 to 12. The Radiation Safety System for the ALS has been designed and built with a primary goal of providing protection against inadvertent personnel exposure to gamma and neutron radiation and, secondarily, to enhance the electrical safety of select magnet power supplies

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

Energy Technology Data Exchange (ETDEWEB)

J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

2014-01-01

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

Advanced gas cooled reactors - Designing for safety

International Nuclear Information System (INIS)

Keen, Barry A.

1990-01-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

Advanced gas cooled reactors - Designing for safety

Energy Technology Data Exchange (ETDEWEB)

Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

1990-07-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

Safety characteristics of the US advanced liquid metal reactor core

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

1991-01-01

The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

System safety engineering in the development of advanced surface transportation vehicles

Science.gov (United States)

Arnzen, H. E.

1971-01-01

Applications of system safety engineering to the development of advanced surface transportation vehicles are described. As a pertinent example, the paper describes a safety engineering efforts tailored to the particular design and test requirements of the Tracked Air Cushion Research Vehicle (TACRV). The test results obtained from this unique research vehicle provide significant design data directly applicable to the development of future tracked air cushion vehicles that will carry passengers in comfort and safety at speeds up to 300 miles per hour.

Advanced vehicle dynamics of heavy trucks with the perspective of road safety

Science.gov (United States)

Trigell, Annika Stensson; Rothhämel, Malte; Pauwelussen, Joop; Kural, Karel

2017-10-01

This paper presents state-of-the art within advanced vehicle dynamics of heavy trucks with the perspective of road safety. The most common accidents with heavy trucks involved are truck against passenger cars. Safety critical situations are for example loss of control (such as rollover and lateral stability) and a majority of these occur during speed when cornering. Other critical situations are avoidance manoeuvre and road edge recovery. The dynamic behaviour of heavy trucks have significant differences compared to passenger cars and as a consequence, successful application of vehicle dynamic functions for enhanced safety of trucks might differ from the functions in passenger cars. Here, the differences between vehicle dynamics of heavy trucks and passenger cars are clarified. Advanced vehicle dynamics solutions with the perspective of road safety of trucks are presented, beginning with the topic vehicle stability, followed by the steering system, the braking system and driver assistance systems that differ in some way from that of passenger cars as well.

Emergency Diesel: Safety-related instrumentation and control with programmable logic controllers

International Nuclear Information System (INIS)

Breidenich, G.; Luedtke, M.

2004-01-01

This report presents a new concept for the design of emergency diesel equipment protection circuits as a part of the safety related instrumentation in the nuclear power plant Biblis, units A and B. The concept was implemented with state of the art SIMATIC S7/316 programmable logic controllers (PLCs) and can be adapted to any system with high availability requirements (e.g. power plant turbines, aircraft engines, mining pumps etc). (orig.)

Advances in and uses of gamma-ray field instrumentation at Los Alamos

International Nuclear Information System (INIS)

Halbig, J.K.; Klosterbuer, S.F.; Russo, P.A.; Sprinkle, J.K. Jr.; Smith, S.E.; Ianakiev, K.

1994-01-01

We are developing a set of tools to be used by the Safeguards Assay Group to solve problems found in safeguards and the domestic nuclear industry. The tools are also applicable to problems dealing with the environment, defense, and other areas of national and international interest. We have used extensively the advances in hardware and software since our last multichannel analyzer (MCA) development activities over a decade ago. We are also using our experience with and feedback from users of our previous instruments. In analyzing the instrument needs of our constituents and the characteristics of our previous instruments, which we think have inhibited their broader use, we have concluded that uses for an MCA-type instrument are widely varied and fundamentally changing,and that any new instruments should include a versatile, widely used hardware interface, which is as independent as possible of hardware standards, and which is readily interfaced to the computers or controllers rapidly evolving in the commercial sector. In addition, software tools must be provided that allow Los Alamos, users, and third parties to quickly and conveniently develop software specific to the user or the measurement to control the basic instrument we develop. This paper deals mainly with a miniature and modular multi-channel analyzer (M 3 CA) and its applications

Studies on environment safety and application of advanced reactor for inland nuclear power plants

International Nuclear Information System (INIS)

Wei, L.; Jie, L.

2014-01-01

To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

The qualification of electrical components and instrumentations relevant to safety

International Nuclear Information System (INIS)

Zambardi, F.

1989-03-01

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and peculiar aspects are discussed and the qualification by testing is taken into special account. Technical and organizational aspects related to a plant qualification program are also focused. The report ends with a look to the most significant research and development activities. (author)

Development of FPGA-based safety-related instrumentation and control systems

Energy Technology Data Exchange (ETDEWEB)

Oda, N.; Tanaka, A.; Izumi, M.; Tarumi, T.; Sato, T. [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

2004-07-01

Toshiba has developed systems which perform signal processing by field programmable gate arrays (FPGA) for safety-related instrumentation and control systems. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing units (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. Considering application to safety-related systems, nonvolatile and non rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. The systems which Toshiba developed this time are Power range Monitor (PRM) and Trip Module (TM). These systems are compatible with the conventional analog-based systems and the CPU-based systems. Therefore, requested cost for upgrading will be minimized. Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)

Enhanced in-pile instrumentation at the advanced test reactor

Energy Technology Data Exchange (ETDEWEB)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

Science.gov (United States)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

New technology for BWR power plant control and instrumentation

International Nuclear Information System (INIS)

Takano, Yoshiyuki; Nakamura, Makoto; Murata, Fumio.

1992-01-01

Nuclear power plants are facing strong demands for higher reliability and cost-performance in their control and instrumentation systems. To meet these needs, Hitachi is developing advanced control and instrumentation technology by rationalizing the conventional technology in that field. The rationalization is done through the utilization of reliable digital technology and optical transmission technology, and others, which are now commonly used in computer applications. The goal of the development work is to ensure safe, stable operation of the plant facilities and to secure harmony between man and machine. To alleviate the burdens of the operators, the latest electronic devices are being employed to create an advanced man-machine interface, and to promote automatic operation of the plant based upon the automatic operation of individual systems. In addition, the control and instrumentation system, including the safety system, incorporates more and more digital components in order to further enhance the reliability and maintainability of the plant. (author)

Development of the Advanced Nuclear Safety Information Management (ANSIM) System

Energy Technology Data Exchange (ETDEWEB)

Sohn, Jae Min; Ko, Young Cheol; Song, Tai Gil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Korea has become a technically independent nuclear country and has grown into an exporter of nuclear technologies. Thus, nuclear facilities are increasing in significance at KAERI (Korea Atomic Energy Research Institute), and it is time to address the nuclear safety. The importance of nuclear safety cannot be overemphasized. Therefore, a management system is needed urgently to manage the safety of nuclear facilities and to enhance the efficiency of nuclear information. We have established ISP (Information Strategy Planning) for the Integrated Information System of nuclear facility and safety management. The purpose of this paper is to develop a management system for nuclear safety. Therefore, we developed the Advanced Nuclear Safety Information Management system (hereinafter referred to as the 'ANSIM system'). The ANSIM system has been designed and implemented to computerize nuclear safety information for standardization, integration, and sharing in real-time. Figure 1 shows the main home page of the ANSIM system. In this paper, we describe the design requirements, contents, configurations, and utilizations of the ANSIM system

Advanced neutron instrumentation at FRM-II

International Nuclear Information System (INIS)

Petry, Winfried

2003-01-01

The construction of the new German high flux neutron source FRM-II is finished and FRM-II is waiting for its licence to start nuclear operation. With the beginning of the routine operation 22 instruments will be in action, including 5 irradiation facilities and 17 beam tube instruments, most of them use neutron scattering techniques. Additional instruments are under construction. Some of these instruments are unique, others are expected to be the best of their kind, all instruments are based on innovative techniques. (author)

The 5th questionnaire report of safety control on instrument in nuclear medicine laboratory

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-08-01

The questionnaire was done every three years from 1986 for the ultimate purpose of safe medical examinations and this 5th one was performed in May, 1998 for the period of April, 1995-March, 1998. Subjects were 1,258 nuclear medicine facilities and answers were obtained in 81.6%. Questionnaire concerned the personnel involved in nuclear medical examinations, instruments, accidents occurred, matters possibly leading to accident, improvement in safety control, serious trouble and breakage of the instrument, request for the instrument manufacturers and so on. Summaries were: numbers of medical radiology technicians were increased, in vitro tests were decreased, SPECT instruments came into wide use, in accident and improvement cases, examination beds were arousing much interest, concerns to examine were further required, communication with the manufacturers was insufficient, and problems for Y2K were pointed out to be resolved. (K.H.)

The advanced neutron source safety approach and plans

International Nuclear Information System (INIS)

Harrington, R.M.

1989-01-01

The Advanced Neutron Source (ANS) is a user facility for all areas of neutron research proposed for construction at the Oak Ridge National Laboratory. The neutron source is planned to be a 350-MW research reactor. The reactor, currently in conceptual design, will belong to the United States Department of Energy (USDOE). The safety approach and planned elements of the safety program for the ANS are described. The safety approach is to incorporate USDOE requirements [which, by reference, include appropriate requirements from the United States Nuclear Regulatory Commission (USNRC) and other national and state regulatory agencies] into the design, and to utilize probabilistic risk assessment (PRA) techniques during design to achieve extremely low probability of severe core damage. The PRA has already begun and will continue throughout the design and construction of the reactor. Computer analyses will be conducted for a complete spectrum of accidental events, from anticipated events to very infrequent occurrences. 8 refs., 2 tabs

The Advanced Neutron Source safety approach and plans

International Nuclear Information System (INIS)

Harrington, R.M.

1990-01-01

The Advanced Neutron Source (ANS) is a user facility proposed for construction at the Oak Ridge National Laboratory for all areas of neutron research. The neutron source is planned to be a 350-MW research reactor. The reactor, currently in conceptual design, will belong to the United States Department of Energy (USDOE). The safety approach and planned elements of the safety program for the ANS are described. The safety approach is to incorporate USDOE requirements (which, by reference, include appropriate requirements from the United States Nuclear Regulatory Commission (USNRC) and other national and state regulatory agencies) into the design, and to utilize probabilistic risk assessment (PRA) techniques during design to achieve extremely low probability of severe core damage. The PRA has already begun and will continue throughout the design and construction of the reactor. Computer analyses will be conducted for a complete spectrum of accidental events, from anticipated events to very infrequent occurrences

Recent advances in the instrumental techniques for the analysis of modern materials (II)

International Nuclear Information System (INIS)

Ahmed, M.

1990-01-01

Inductively Coupled Plasma Mass Spectrometry ICP-MS a logical development of equally established sister technique of ICP-AEA discussed in part-1 of this series of article on modern analytical techniques. The rapid adaptation of argon plasma as ion source for time of flight quadrupole mass analyser has led to the development of truly integrated instrumental technique for analysis of solutions and slurries. The powerful combination with laser ablation device has made the direct analysis of geological, geochemical and other complex conducting and non conducting samples possible in days rather months at sub ppm levels. Parallel development in computer hardware and software has made the instrumental optimization easy enabling the generation of meaningful analytical data a matter of routine. The limitations imposed by spectroscopic and non restricted the variety of matrices and materials covered by ICP-MS of LA-ICP-MS. The technique has provided it formidable analytical power in wide areas of industrial environmental, social, biological and break through advanced materials used in space mass communication, transportation and general areas of advanced analytical chemistry. It is expected that in combination with other instrumental methods as HPLC, ETC, ion chromatography. ICP-MS shall continue to dominate well into the 21st century. (author)

Luxury cruise? The safety potential of advanced cruise control.

NARCIS (Netherlands)

Oei, H.L.

2003-01-01

The principles of advanced cruise control (ACC) are outlined and the requirements for an ACC system are described. An intelligent cruise control system fitted in a Nissan Primera was tested on the road over a 2-week period by 10 drivers, eight of which were experts in road safety. Most test-drives

Participant Assessments of Aviation Safety Inspector Training for Technically Advanced Aircraft

National Research Council Canada - National Science Library

Chidester, Thomas; Hackworth, Carla; Knecht, William

2007-01-01

.... Currently, Federal Aviation Administration (FAA) aviation safety inspectors are required to inspect technically advanced aircraft, check certified flight instructors, and conduct surveillance of designated pilot examiners who are certifying pilots...

Application of a structural model for advanced analysis in the evaluation of nuclear safety

International Nuclear Information System (INIS)

Landesmann, Alexandre; Barros, Francisco Claudio Pereira de; Batista, Eduardo de Miranda

2003-01-01

The Advanced Analysis concept, which means the direct consideration of both physical and geometric nonlinear effects in the analysis and design of steel buildings structures, represents the state-of-art in the field of structural analysis by this beginning of the 21 st century. In this context, the present paper presents an Advanced Analysis methodology applied to the Safety Evaluation of high hazardous civil structures. This Safety Evaluation plays an important part in the regulators position as a step in the licensing process performed by CNEN - Brazilian Nuclear Energy Commission. The proposed Advance Analysis procedure is implemented by a refined second-order plastic hinge model. The application of this model allows to carry out: the description of the inelastic structural behavior; the identification of the collapse mechanism; the ultimate load level; structural safety's level and the service ability limit. (author)

1st Advanced Marine Renewable Energy Instrumentation Experts Workshop: April 5-7, 2011

International Nuclear Information System (INIS)

2011-01-01

The U.S. marine energy industry is actively pursuing development of offshore wind and marine hydrokinetic (MHK) energy systems. Experience in the wind energy sector demonstrates that new technology development requires thorough measurement and characterization of the environmental conditions prevalent at installation sites and of technology operating in the field. Presently, there are no turn-key instrumentation system solutions that meet the measurement needs of the marine energy industry. The 1st Advanced Marine Renewable Energy Instrumentation Experts Workshop brought together technical experts from government laboratories, academia, and industry representatives from marine energy, wind, offshore oil and gas, and instrumentation developers to present and discuss the instrumentation needs of the marine energy industry. The goals of the meeting were to: (1) Share the latest relevant knowledge among technical experts; (2) Review relevant state-of-the-art field measurement technologies and methods; (3) Review lessons learned from recent field deployments; (4) Identify synergies across different industries; (5) Identify gaps between existing and needed instrumentation capabilities; (6) Understand who are the leading experts; (7) Provide a forum where stakeholders from the marine energy industry could provide substantive input in the development of new marine energy field deployable instrumentation packages.

1st Advanced Marine Renewable Energy Instrumentation Experts Workshop: April 5-7, 2011

Energy Technology Data Exchange (ETDEWEB)

2011-10-01

The U.S. marine energy industry is actively pursuing development of offshore wind and marine hydrokinetic (MHK) energy systems. Experience in the wind energy sector demonstrates that new technology development requires thorough measurement and characterization of the environmental conditions prevalent at installation sites and of technology operating in the field. Presently, there are no turn-key instrumentation system solutions that meet the measurement needs of the marine energy industry. The 1st Advanced Marine Renewable Energy Instrumentation Experts Workshop brought together technical experts from government laboratories, academia, and industry representatives from marine energy, wind, offshore oil and gas, and instrumentation developers to present and discuss the instrumentation needs of the marine energy industry. The goals of the meeting were to: (1) Share the latest relevant knowledge among technical experts; (2) Review relevant state-of-the-art field measurement technologies and methods; (3) Review lessons learned from recent field deployments; (4) Identify synergies across different industries; (5) Identify gaps between existing and needed instrumentation capabilities; (6) Understand who are the leading experts; (7) Provide a forum where stakeholders from the marine energy industry could provide substantive input in the development of new marine energy field deployable instrumentation packages.

Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

International Nuclear Information System (INIS)

Ridha Roslan; Nik Mohd Faiz Khairuddin

2013-01-01

Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

2014-08-15

The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

Enhanced operational safety of BWRs by advanced computer technology and human engineering

International Nuclear Information System (INIS)

Tomizawa, T.; Fukumoto, A.; Neda, T.; Toda, Y.; Takizawa, Y.

1984-01-01

In BWR nuclear power plants, where unit capacity is increasing and the demand for assured safety is growing, it has become important for the information interface between man and machine to work smoothly. Efforts to improve man-machine communication have been going on for the past ten years in Japan. Computer facilities and colour CRT display systems are amongst the most useful new methods. Advanced computer technology has been applied to operating plants and found to be very helpful for safe operation. A display monitoring system (DMS) is in operation in a 1100 MW(e) BWR plant. A total combination test was successfully completed on the 'plant operation by displayed information and automation' system (PODIA) in February 1983 before shipment to the site. The objective of this test was to verify the improved qualification of the newly developed advanced PODIA man-machine system by this enlarged fabrication test concept. In addition, the development of special graphics displays for the main control room and technical support centre to assist operators in assessing plant safety and diagnosing problems is required to meet post-TMI regulations. For this purpose, a prototype safety parameter display system (called Toshiba SPDS) with two colour CRT displays and a computer (TOSBAC-7/70) was developed in 1981 as an independent safety monitoring system. The PODIA and SPDS are now independent systems, but their combination has been found to be more useful and valuable for nuclear power plant safety. The paper discusses supervisory and operational concepts in the advanced main control room including SPDS, and describes the PODIA and SPDS verification tests including the valuable experience obtained after improvements in the qualification of these systems had been made to satisfactory operational safety levels. (author)

Safety aspects of advanced fuels irradiations in EBR-II

International Nuclear Information System (INIS)

Lehto, W.K.

1975-09-01

Basic safety questions such as MFCI, loss-of-Na bond, pin behavior during design basis transients, and failure propagation were evaluated as they pertain to advanced fuels in EBR-II. With the exception of pin response to the unlikely loss-of-flow transient, the study indicates that irradiation of significant numbers of advanced fueled subassemblies in EBR-II should pose no safety problems. The analysis predicts, however, that Na boiling may occur during the postulated design basis unlikely loss-of-flow transient in subassemblies containing He-bonded fuel pins with the larger fuel-clad gaps. The calculations indicate that coolant temperatures at top of core in the limiting S/A's, containing the He bonded pins, would reach approximately 1480 0 F during the transient without application of uncertainty factors. Inclusion of uncertainties could result in temperature predictions which approach coolant boiling temperatures (1640 0 F). Further analysis of He-bonded pins is being done in this potential problem area, e.g., to apply best estimates of uncertainty factors and to determine the sensitivity of the preliminary results to gap conductance

Recent advances in radiation protection instrumentation

International Nuclear Information System (INIS)

Babu, D.A.R.

2012-01-01

Radiation protection instrumentation plays very important role in radiation protection and surveillance programme. Radiation detector, which appears at the frontal end of the instrument, is an essential component of these instruments. The instrumental requirement of protection level radiation monitoring is different from conventional radiation measuring instruments. Present paper discusses the new type of nuclear radiation detectors, new protection level instruments and associated electronic modules for various applications. Occupational exposure to ionizing radiation can occur in a range of industries, such as nuclear power plants; mining and milling; medical institutions; educational and research establishments; and nuclear fuel cycle facilities. Adequate radiation protection to workers is essential for the safe and acceptable use of radioactive materials for different applications. The radiation exposures to the individual radiation workers and records of their cumulative radiation doses need to be routinely monitored and recorded

Inspector-instrument interface in portable NDA instrumentation

International Nuclear Information System (INIS)

Halbig, J.K.; Klosterbuer, S.F.

1981-01-01

Recent electronics technology advances make it possible to design sophisticated instruments in small packages for convenient field implementation. An inspector-instrument interface design that allows communication of procedures, responses, and results between the instrument and user is presented. This capability has been incorporated into new spent-fuel instrumentation and a battery-powered multichannel analyzer

Inspector-instrument interface in portable NDA instrumentation

International Nuclear Information System (INIS)

Halbig, J.K.; Klosterbuer, S.F.

1981-01-01

Recent electronics technology advances make it possible to design sophisticated instruments in small packages for convenient field implementation. This report describes an inspector-instrument interface design which allows communication of procedures, responses, and results between the instrument and user. The interface has been incorporated into new spent-fuel instrumentation and a battery-powered multichannel analyzer

Advanced power reactors with improved safety characteristics

International Nuclear Information System (INIS)

Birkhofer, A.

1994-01-01

The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

Launching the dialogue: Safety and innovation as partners for success in advanced manufacturing.

Science.gov (United States)

Geraci, C L; Tinkle, S S; Brenner, S A; Hodson, L L; Pomeroy-Carter, C A; Neu-Baker, N

2018-06-01

Emerging and novel technologies, materials, and information integrated into increasingly automated and networked manufacturing processes or into traditional manufacturing settings are enhancing the efficiency and productivity of manufacturing. Globally, there is a move toward a new era in manufacturing that is characterized by: (1) the ability to create and deliver more complex designs of products; (2) the creation and use of materials with new properties that meet a design need; (3) the employment of new technologies, such as additive and digital techniques that improve on conventional manufacturing processes; and (4) a compression of the time from initial design concept to the creation of a final product. Globally, this movement has many names, but "advanced manufacturing" has become the shorthand for this complex integration of material and technology elements that enable new ways to manufacture existing products, as well as new products emerging from new technologies and new design methods. As the breadth of activities associated with advanced manufacturing suggests, there is no single advanced manufacturing industry. Instead, aspects of advanced manufacturing can be identified across a diverse set of business sectors that use manufacturing technologies, ranging from the semiconductors and electronics to the automotive and pharmaceutical industries. The breadth and diversity of advanced manufacturing may change the occupational and environmental risk profile, challenge the basic elements of comprehensive health and safety (material, process, worker, environment, product, and general public health and safety), and provide an opportunity for development and dissemination of occupational and environmental health and safety (OEHS) guidance and best practices. It is unknown how much the risk profile of different elements of OEHS will change, thus requiring an evolution of health and safety practices. These changes may be accomplished most effectively through multi

Advanced Technologies For Heterodyne Radio Astronomy Instrumentation - Part1 By A. Pavolotsky, And Advanced Technologies For Heterodyne Radio Astronomy Instrumentation - Part2 By V. Desmaris

Science.gov (United States)

Pavolotsky, Alexey

2018-01-01

Modern and future heterodyne radio astronomy instrumentation critically depends on availability of advanced fabrication technologies and components. In Part1 of the Poster, we present the thin film fabrication process for SIS mixer receivers, utilizing either AlOx, or AlN barrier superconducting tunnel junctions developed and supported by GARD. The summary of the process design rules is presented. It is well known that performance of waveguide mixer components critically depends on accuracy of their geometrical dimensions. At GARD, all critical mechanical parts are 3D-mapped with a sub-um accuracy. Further progress of heterodyne instrumentation requires new efficient and compact sources of LO signal. We present SIS-based frequency multiplier, which could become a new option for LO source. Future radio astronomy THz receivers will need waveguide components, which fabricating due to their tiny dimensions is not feasible by traditional mechanical machining. We present the alternative micromachining technique for fabricating waveguide component for up 5 THz band and probably beyond.

ATLAS Facility and Instrumentation Description Report

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Moon, Sang Ki; Park, Hyun Sik

2009-06-01

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). The ATLAS is a half-height and 1/288-volume scaled test facility with respect to the APR1400. The fluid system of the ATLAS consists of a primary system, a secondary system, a safety injection system, a break simulating system, a containment simulating system, and auxiliary systems. The primary system includes a reactor vessel, two hot legs, four cold legs, a pressurizer, four reactor coolant pumps, and two steam generators. The secondary system of the ATLAS is simplified to be of a circulating looptype. Most of the safety injection features of the APR1400 and the OPR1000 are incorporated into the safety injection system of the ATLAS. In the ATLAS test facility, about 1300 instrumentations are installed to precisely investigate the thermal-hydraulic behavior in simulation of the various test scenarios. This report describes the scaling methodology, the geometric data of the individual component, and the specification and the location of the instrumentations which are specific to the simulation of 50% DVI line break accident of the APR1400 for supporting the 50 th OECD/NEA International Standard Problem Exercise (ISP-50)

A toolbox for safety instrumented system evaluation based on improved continuous-time Markov chain

Science.gov (United States)

Wardana, Awang N. I.; Kurniady, Rahman; Pambudi, Galih; Purnama, Jaka; Suryopratomo, Kutut

2017-08-01

Safety instrumented system (SIS) is designed to restore a plant into a safe condition when pre-hazardous event is occur. It has a vital role especially in process industries. A SIS shall be meet with safety requirement specifications. To confirm it, SIS shall be evaluated. Typically, the evaluation is calculated by hand. This paper presents a toolbox for SIS evaluation. It is developed based on improved continuous-time Markov chain. The toolbox supports to detailed approach of evaluation. This paper also illustrates an industrial application of the toolbox to evaluate arch burner safety system of primary reformer. The results of the case study demonstrates that the toolbox can be used to evaluate industrial SIS in detail and to plan the maintenance strategy.

Advancing perinatal patient safety through application of safety science principles using health IT.

Science.gov (United States)

Webb, Jennifer; Sorensen, Asta; Sommerness, Samantha; Lasater, Beth; Mistry, Kamila; Kahwati, Leila

2017-12-19

The use of health information technology (IT) has been shown to promote patient safety in Labor and Delivery (L&D) units. The use of health IT to apply safety science principles (e.g., standardization) to L&D unit processes may further advance perinatal safety. Semi-structured interviews were conducted with L&D units participating in the Agency for Healthcare Research and Quality's (AHRQ's) Safety Program for Perinatal Care (SPPC) to assess units' experience with program implementation. Analysis of interview transcripts was used to characterize the process and experience of using health IT for applying safety science principles to L&D unit processes. Forty-six L&D units from 10 states completed participation in SPPC program implementation; thirty-two (70%) reported the use of health IT as an enabling strategy for their local implementation. Health IT was used to improve standardization of processes, use of independent checks, and to facilitate learning from defects. L&D units standardized care processes through use of electronic health record (EHR)-based order sets and use of smart pumps and other technology to improve medication safety. Units also standardized EHR documentation, particularly related to electronic fetal monitoring (EFM) and shoulder dystocia. Cognitive aids and tools were integrated into EHR and care workflows to create independent checks such as checklists, risk assessments, and communication handoff tools. Units also used data from EHRs to monitor processes of care to learn from defects. Units experienced several challenges incorporating health IT, including obtaining organization approval, working with their busy IT departments, and retrieving standardized data from health IT systems. Use of health IT played an integral part in the planning and implementation of SPPC for participating L&D units. Use of health IT is an encouraging approach for incorporating safety science principles into care to improve perinatal safety and should be incorporated

Advancements in the design of safety-related systems and components of the MARS nuclear plant

International Nuclear Information System (INIS)

Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

1992-01-01

In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

Advanced analysis and design for fire safety of steel structures

CERN Document Server

Li, Guoqiang

2013-01-01

Advanced Analysis and Design for Fire Safety of Steel Structures systematically presents the latest findings on behaviours of steel structural components in a fire, such as the catenary actions of restrained steel beams, the design methods for restrained steel columns, and the membrane actions of concrete floor slabs with steel decks. Using a systematic description of structural fire safety engineering principles, the authors illustrate the important difference between behaviours of an isolated structural element and the restrained component in a complete structure under fire conditions. The book will be an essential resource for structural engineers who wish to improve their understanding of steel buildings exposed to fires. It is also an ideal textbook for introductory courses in fire safety for master’s degree programs in structural engineering, and is excellent reading material for final-year undergraduate students in civil engineering and fire safety engineering. Furthermore, it successfully bridges th...

Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2014

Energy Technology Data Exchange (ETDEWEB)

Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-07-01

The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

Instrumentation for mine safety: fire and smoke problems and solutions

International Nuclear Information System (INIS)

Stevens, R.B.

1982-01-01

Underground fires continue to be one of the most serious hazards to life and property in the mining industry. Although underground mines are analogous to high-rise buildings where persons are isolated from immediate escape or rescue, application of technology to locate and control fire hazards while still in their controllable state is slow to be implemented in underground mines. This paper describes several USBM (Bureau of Mines) safety programs which included in-mine testing with mine fire and smoke sensors, telemetry and instrumentation to develop recommendations for improving mine fire safety. It is hoped that the technology developed during these programs can be added to other programs to provide the mining industry with the necessary fire safety facts. By recognizing fire potentials and being provided with cost-effective, proven components that will perform reliably under the poor environmental conditions of mining, mine operators can provide protection for their working life and property equal to that which they provide for themselves and their families at home. The basis of this report is two USBM programs for fire protection in metal and nonmetal mines and one coal program. The data was collected beginning in May 1974 and continuing through the present with underground tests of a South African fire system installed at Magma Mine in Superior, Arizona, and a computer-assisted, experimental system at Peabody Coal Mine in Pawnee, Illinois

An investigation of fluid mixing with safety injection in advanced reactors

International Nuclear Information System (INIS)

Cha, Jong Hee; Won, Soon Yean; Chung, Moon Ki; Jun, Hyung Gil

1994-01-01

The objective of this work is to investigate the fluid mixing phenomena in aspect of pressurized thermal shock(PTS) in an advanced PWR vessel downcomer during transient cooldown with safety injection. It provides comparison of fluid mixing characteristics between AP 600 DVI, designed by Westinghouse, and ABB CE System 80+ DVI, and the effects of deflector at the reactor downcomer. In order to investigate the fluid mixing phenomena in the downcomer of an advanced PWR, the flow visualization tests and the salt concentration tests were conducted in a 1/7-scale acrylic transparent model, which was designed and built based on AP 600 reactor geometry. The behaviour of the safety injection flow in downcomer associated with mixing phenomenon can be observed during visualization test, and time-dependent mixing rate between safety injection fluid and existing coolant can be determined with concentration test. Visualization tests were performed by the dye injection method. The results of concentration measurements were compared with the calculation using the REMIX code. During the tests, difference between AP 600 DVI flow and ABB CE System 80+ DVI flow and the effect of the deflector were observed

Reliability analysis and computation of computer-based safety instrumentation and control used in German nuclear power plant. Final report

International Nuclear Information System (INIS)

Ding, Yongjian; Krause, Ulrich; Gu, Chunlei

2014-01-01

The trend of technological advancement in the field of safety instrumentation and control (I and C) leads to increasingly frequent use of computer-based (digital) control systems which consisting of distributed, connected bus communications computers and their functionalities are freely programmable by qualified software. The advantages of the new I and C system over the old I and C system with hard-wired technology are e.g. in the higher flexibility, cost-effective procurement of spare parts, higher hardware reliability (through higher integration density, intelligent self-monitoring mechanisms, etc.). On the other hand, skeptics see the new technology with the computer-based I and C a higher potential by influences of common cause failures (CCF), and the easier manipulation by sabotage (IT Security). In this joint research project funded by the Federal Ministry for Economical Affaires and Energy (BMWi) (2011-2014, FJZ 1501405) the Otto-von-Guericke-University Magdeburg and Magdeburg-Stendal University of Applied Sciences are therefore trying to develop suitable methods for the demonstration of the reliability of the new instrumentation and control systems with the focus on the investigation of CCF. This expertise of both houses shall be extended to this area and a scientific contribution to the sound reliability judgments of the digital safety I and C in domestic and foreign nuclear power plants. First, the state of science and technology will be worked out through the study of national and international standards in the field of functional safety of electrical and I and C systems and accompanying literature. On the basis of the existing nuclear Standards the deterministic requirements on the structure of the new digital I and C system will be determined. The possible methods of reliability modeling will be analyzed and compared. A suitable method called multi class binomial failure rate (MCFBR) which was successfully used in safety valve applications will be

JACoW Safety instrumented systems and the AWAKE plasma control as a use case

CERN Document Server

Blanco Viñuela, Enrique; Fernández Adiego, Borja; Speroni, Roberto

2018-01-01

Safety is likely the most critical concern in many process industries, yet there is a general uncertainty on the proper engineering to reduce the risks and ensure the safety of persons or material at the same time as providing the process control system. Some of the reasons for this misperception are unclear requirements, lack of functional safety engineering knowledge or incorrect protection functionalities attributed to the BPCS (Basic Process Control System). Occasionally the control engineers are not aware of the hazards inherent to an industrial process and this causes an incorrect design of the overall controls. This paper illustrates the engineering of the SIS (Safety Instrumented System) and the BPCS of the plasma vapour controls of the AWAKE R&D; project, the first proton-driven plasma wakefield acceleration experiment in the world. The controls design and implementation refers to the IEC61511/ISA84 standard, including technological choices, design, operation and maintenance. Finally, the publica...

Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

International Nuclear Information System (INIS)

2009-01-01

In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

Qualification issues for advanced light-water reactor protection systems

International Nuclear Information System (INIS)

Korsah, K.; Clark, R.L.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

Proof-testing strategies induced by dangerous detected failures of safety-instrumented systems

International Nuclear Information System (INIS)

Liu, Yiliu; Rausand, Marvin

2016-01-01

Some dangerous failures of safety-instrumented systems (SISs) are detected almost immediately by diagnostic self-testing as dangerous detected (DD) failures, whereas other dangerous failures can only be detected by proof-testing, and are therefore called dangerous undetected (DU) failures. Some items may have a DU- and a DD-failure at the same time. After the repair of a DD-failure is completed, the maintenance team has two options: to perform an insert proof test for DU-failure or not. If an insert proof test is performed, it is necessary to decide whether the next scheduled proof test should be postponed or performed at the scheduled time. This paper analyzes the effects of different testing strategies on the safety performance of a single channel of a SIS. The safety performance is analyzed by Petri nets and by approximation formulas and the results obtained by the two approaches are compared. It is shown that insert testing improves the safety performance of the channel, but the feasibility and cost of the strategy may be a hindrance to recommend insert testing. - Highlights: • Identify the tests induced by detected failures. • Model the testing strategies following DD-failures. • Propose analytical formulas for effects of strategies. • Simulate and verify the proposed models.

Analysis on Configuration of I and C Systems for an Advanced HANARO Reactors

International Nuclear Information System (INIS)

Park, Gee Yong; Jung, H. S.; Ryu, J. S.; Park, C.

2006-01-01

In an advanced HANARO reactor (AHR), the instrumentation and control (I and C) systems are designed based on the digital system rather than the analog system installed in an existing HANARO instrumentation and control systems. While the safety and functionality of analog-based instrumentation and control system are experienced over a long period of operating time and also well-validated, the obsolescence and the lack of flexibility of this system have to move from the analog technology to the digital technology in the instrumentation and control systems to be used in nuclear power plants as well as nuclear research reactors. For establishing the adequate structure of instrumentation and control systems for an AHR, various instrumentation and control architectures are analyzed for their merits and demerits for use in I and C systems of an AHR and the most promising instrumentation and control architecture for an AHR are drawn from this analysis. The conceptual configuration of a digital-based safety shutdown system is proposed in this report

RBMK nuclear reactors: Proposals for instrumentation and control improvements to enhanced safety and availability. IEC technical report of type 3. Working material

International Nuclear Information System (INIS)

1995-01-01

The present material presents a CD+V draft report ''RBMK nuclear reactors: Proposals for instrumentation and control improvements to enhance safety and availability'' prepared by the Joint IEC/IAEA team during 1993-1995. Experience has demonstrated the need to improve the safety instrumentation of the RBMK type reactors using well proven modern technology. The working group identified the upgrades and changes of the highest priority based on the evaluation of the RBMK systems and the events where the instrumentation was found to be inadequate for safe operation. The subjects discussed in this document were not selected on a systematic basis but were selected by the IEC and IAEA experts as considered to be appropriate to the activities of the IEC and for which technical experience was available. The items identified therefore do not reflect any ranking of the safety issues or any priority or impact on safety of any of the measures were they to be implemented. Many important safety issued and areas where physical measures are required to improve safety have been omitted and indeed not even acknowledged in this document. The recommendations presented in the document differ from those normally produced by the IEC in the form of standards as they are of a transitory nature and some have already been overtaken by the continuing process of improvements to plant safety. Figs and tabs

The strategy for intelligent integrated instrumentation and control system development

International Nuclear Information System (INIS)

Kwon, Kee Choon; Ham, Chang Shik

1995-01-01

All of the nuclear power plants in Korea are operating with analog instrumentation and control ( I and C) equipment which are increasingly faced with frequent troubles, obsolescence and high maintenance expenses. Electrical and computer technology has improved rapidly in recent years and has been applied to other industries. So it is strongly recommended we adopt modern digital and computer technology to improve plant safety and availability. The advanced I and C system, namely, Integrated Intelligent Instrumentation and Control System (I 3 Cs) will be developed for beyond the next generation nuclear power plant. I 3 CS consists of three major parts, the advanced compact workstation, distributed digital control and protection system including Automatic Start-up/Shutdown Intelligent Control System (ASICS) and the computer-based alarm processing and operator support system, namely, Diagnosis, Response, and operator Aid Management System (DREAMS)

Instrumentation, Control, and Intelligent Systems

International Nuclear Information System (INIS)

Not Available

2005-01-01

Abundant and affordable energy is required for U.S. economic stability and national security. Advanced nuclear power plants offer the best near-term potential to generate abundant, affordable, and sustainable electricity and hydrogen without appreciable generation of greenhouse gases. To that end, Idaho National Laboratory (INL) has been charged with leading the revitalization of nuclear power in the U.S. The INL vision is to become the preeminent nuclear energy laboratory with synergistic, world-class, multi-program capabilities and partnerships by 2015. The vision focuses on four essential destinations: (1) Be the preeminent internationally-recognized nuclear energy research, development, and demonstration laboratory; (2) Be a major center for national security technology development and demonstration; (3) Be a multi-program national laboratory with world-class capabilities; (4) Foster academic, industry, government, and international collaborations to produce the needed investment, programs, and expertise. Crucial to that effort is the inclusion of research in advanced instrumentation, control, and intelligent systems (ICIS) for use in current and advanced power and energy security systems to enable increased performance, reliability, security, and safety. For nuclear energy plants, ICIS will extend the lifetime of power plant systems, increase performance and power output, and ensure reliable operation within the system's safety margin; for national security applications, ICIS will enable increased protection of our nation's critical infrastructure. In general, ICIS will cost-effectively increase performance for all energy security systems

Advanced Range Safety System for High Energy Vehicles

Science.gov (United States)

Claxton, Jeffrey S.; Linton, Donald F.

2002-01-01

The advanced range safety system project is a collaboration between the National Aeronautics and Space Administration and the United States Air Force to develop systems that would reduce costs and schedule for safety approval for new classes of unmanned high-energy vehicles. The mission-planning feature for this system would yield flight profiles that satisfy the mission requirements for the user while providing an increased quality of risk assessment, enhancing public safety. By improving the speed and accuracy of predicting risks to the public, mission planners would be able to expand flight envelopes significantly. Once in place, this system is expected to offer the flexibility of handling real-time risk management for the high-energy capabilities of hypersonic vehicles including autonomous return-from-orbit vehicles and extended flight profiles over land. Users of this system would include mission planners of Space Launch Initiative vehicles, space planes, and other high-energy vehicles. The real-time features of the system could make extended flight of a malfunctioning vehicle possible, in lieu of an immediate terminate decision. With this improved capability, the user would have more time for anomaly resolution and potential recovery of a malfunctioning vehicle.

Cost-effective instrumentation and control upgrades for commercial nuclear power plants surety principles developed at Sandia National Laboratories

International Nuclear Information System (INIS)

Rochau, G.E.; Dalton, L.J.

1998-01-01

Many nuclear power plants use instrument and control systems based on analog electronics. The state of the art in process control and instrumentation has advanced to use digital electronics and incorporate advanced technology. This technology includes: distributed microprocessors, fiber optics, intelligent systems (neutral networks), and advanced displays. The technology is used to optimize processes and enhance the man-machine interface while maintaining control and safety of the processes. Nuclear power plant operators have been hesitant to install this technology because of the cost and uncertainty in the regulatory process. This technology can be directly applied in an operating nuclear power plant provided a surety principle-based 'administrator' hardware system is included in parallel with the upgrade. Sandia National Laboratories has developed a rigorous approach to High Consequence System Surety (HCSS). This approach addresses the key issues of safety, security, and control while satisfying requirements for reliability and quality. We believe that HCSS principles can be applied to nuclear power plants in a manner that allows the off-the-shelf use of process control instrumentation while maintaining a high level of safety and enhancing the plant performance. We propose that an HCSS Administrator be constructed as a standardized approach to address regulatory issues. Such an administrator would allow a plant control system to be constructed with commercially available, state-to-the-art equipment and be customized to the needs of the individual plant operator. (author)

Technological Advances in Cardiovascular Safety Assessment Decrease Preclinical Animal Use and Improve Clinical Relevance.

Science.gov (United States)

Berridge, Brian R; Schultze, A Eric; Heyen, Jon R; Searfoss, George H; Sarazan, R Dustan

2016-12-01

Cardiovascular (CV) safety liabilities are significant concerns for drug developers and preclinical animal studies are predominately where those liabilities are characterized before patient exposures. Steady progress in technology and laboratory capabilities is enabling a more refined and informative use of animals in those studies. The application of surgically implantable and telemetered instrumentation in the acute assessment of drug effects on CV function has significantly improved historical approaches that involved anesthetized or restrained animals. More chronically instrumented animals and application of common clinical imaging assessments like echocardiography and MRI extend functional and in-life structural assessments into the repeat-dose setting. A growing portfolio of circulating CV biomarkers is allowing longitudinal and repeated measures of cardiac and vascular injury and dysfunction better informing an understanding of temporal pathogenesis and allowing earlier detection of undesirable effects. In vitro modeling systems of the past were limited by their lack of biological relevance to the in vivo human condition. Advances in stem cell technology and more complex in vitro modeling platforms are quickly creating more opportunity to supplant animals in our earliest assessments for liabilities. Continuing improvement in our capabilities in both animal and nonanimal modeling should support a steady decrease in animal use for primary liability identification and optimize the translational relevance of the animal studies we continue to do. © The Author 2016. Published by Oxford University Press on behalf of the Institute for Laboratory Animal Research. All rights reserved. For permissions, please email: journals.permissions@oup.com.

Twenty-third water reactor safety information meeting: Volume 2, Human factors research; Advanced I and C hardware and software; Severe accident research; Probabilistic risk assessment topics; Individual plant examination: Proceedings

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1996-03-01

This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 2, present topics in human factors research, advanced instrumentation and control hardware and software, severe accident research, probabilistic risk assessment, and individual plant examination. Individual papers have been cataloged separately.

Editorial: Advances in healthcare provider and patient training to improve the quality and safety of patient care

Directory of Open Access Journals (Sweden)

Elizabeth M. Borycki

2015-09-01

Full Text Available This special issue of the Knowledge Management & E-Learning: An International Journal is dedicated to describing “Advances in Healthcare Provider and Patient Training to Improve the Quality and Safety of Patient Care.” Patient safety is an important and fundamental requirement of ensuring the quality of patient care. Training and education has been identified as a key to improving healthcare provider patient safety competencies especially when working with new technologies such as electronic health records and mobile health applications. Such technologies can be harnessed to improve patient safety; however, if not used properly they can negatively impact on patient safety. In this issue we focus on advances in training that can improve patient safety and the optimal use of new technologies in healthcare. For example, use of clinical simulations and online computer based training can be employed both to facilitate learning about new clinical discoveries as well as to integrate technology into day to day healthcare practices. In this issue we are publishing papers that describe advances in healthcare provider and patient training to improve patient safety as it relates to the use of educational technologies, health information technology and on-line health resources. In addition, in the special issue we describe new approaches to training and patient safety including, online communities, clinical simulations, on-the-job training, computer based training and health information systems that educate about and support safer patient care in real-time (i.e. when health professionals are providing care to patients. These educational and technological initiatives can be aimed at health professionals (i.e. students and those who are currently working in the field. The outcomes of this work are significant as they lead to safer care for patients and their family members. The issue has both theoretical and applied papers that describe advances in patient

Proving autonomous vehicle and advanced driver assistance systems safety : final research report.

Science.gov (United States)

2016-02-15

The main objective of this project was to provide technology for answering : crucial safety and correctness questions about verification of autonomous : vehicle and advanced driver assistance systems based on logic. : In synergistic activities, we ha...

RECENT BEAM MEASUREMENTS AND NEW INSTRUMENTATION AT THE ADVANCED LIGHT SOURCE

International Nuclear Information System (INIS)

Sannibale, Fernando; Baptiste, Kenneth; Barry, Walter; Chin, Michael; Filippetto, Daniele; Jaegerhofer, Lukas; Julian, James; Kwiatkowski, Slawomir; Low, Raymond; Plate, David; Portmann, Gregory; Robin, David; Scarvie, Tomas; Stupakov, Gennady; Weber, Jonah; Zolotorev, Max

2008-01-01

The Advanced Light Source (ALS) in Berkeley was the first of the soft x-ray third generation light source ever built, and since 1993 has been in continuous and successful operation serving a large community of users in the VUV and soft x-ray community. During these years the storage ring underwent through several important upgrades that allowed maintaining the performance of this veteran facility at the forefront. The ALS beam diagnostics and instrumentation have followed a similar path of innovation and upgrade and nowadays include most of the modem and last generation devices and technologies that are commercially available and used in the recently constructed third generation light sources. In this paper we will not focus on such already widely known systems, but we will concentrate our effort in the description of some measurements techniques, instrumentation and diagnostic systems specifically developed at the ALS and used during the last few years

Nuclear electronic instrumentation

International Nuclear Information System (INIS)

Ramirez J, F. J.

2010-01-01

The activities carried out in the Instituto Nacional de Investigaciones Nucleares (ININ) in the field of the nuclear electronic instrumentation included those activities corresponding to the design and production of nuclear instruments in a first stage, as well as the internal activities of design, repair and maintenance that have supported to other projects of the institution during many years. It is mentioned of the presence and constant collaboration of the ININ with the IAEA in different projects and programs. Also, it is mentioned on the establishment of the Radiation Detectors Laboratory, which for their characteristics and repair capacities of radiation detectors of cooled semiconductor, it is only in their specialty. It is emphasized the investigation and the development in the field of new radiation detectors and applications, as well as the important contribution in this field, in institutions like: Mexican Petroleum, National Commission of Nuclear Safety and Safeguards and Federal Commission of Electricity. Finally a position of the future of these activities is made, considering the speed of the advances of the electronic and nuclear technology. (Author)

The integrated code system CASCADE-3D for advanced core design and safety analysis

International Nuclear Information System (INIS)

Neufert, A.; Van de Velde, A.

1999-01-01

The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

Advanced instrumentation for Solar System gravitational physics

Science.gov (United States)

Peron, Roberto; Bellettini, G.; Berardi, S.; Boni, A.; Cantone, C.; Coradini, A.; Currie, D. G.; Dell'Agnello, S.; Delle Monache, G. O.; Fiorenza, E.; Garattini, M.; Iafolla, V.; Intaglietta, N.; Lefevre, C.; Lops, C.; March, R.; Martini, M.; Nozzoli, S.; Patrizi, G.; Porcelli, L.; Reale, A.; Santoli, F.; Tauraso, R.; Vittori, R.

2010-05-01

The Solar System is a complex laboratory for testing gravitational physics. Indeed, its scale and hierarchical structure make possible a wide range of tests for gravitational theories, studying the motion of both natural and artificial objects. The usual methodology makes use of tracking information related to the bodies, fitted by a suitable dynamical model. Different equations of motion are provided by different theories, which can be therefore tested and compared. Future exploration scenarios show the possibility of placing deep-space probes near the Sun or in outer Solar System, thereby extending the available experimental data sets. In particular, the Earth-Moon is the most accurately known gravitational three-body laboratory, which is undergoing a new, strong wave of research and exploration (both robotic and manned). In addition, the benefits of a synergetic study of planetary science and gravitational physics are of the greatest importance (as shown by the success of the Apollo program), especially in the Earth-Moon, Mars-Phobos, Jovian and Saturnian sub-suystems. This scenarios open critical issues regarding the quality of the available dynamical models, i.e. their capability of fitting data without an excessive number of empirical hypotheses. A typical case is represented by the non-gravitational phenomena, which in general are difficult to model. More generally, gravitation tests with Lunar Laser Ranging, inner or outer Solar System probes and the appearance of the so-called 'anomalies'(like the one indicated by the Pioneers), whatever their real origin (either instrumental effects or due to new physics), show the necessity of a coordinated improvement of tracking and modelization techniques. A common research path will be discussed, employing the development and use of advanced instrumentation to cope with current limitations of Solar System gravitational tests. In particular, the use of high-sensitivity accelerometers, combined with microwave and laser

Advances in Field Deployable Instrumented Particles for the Study of Alluvial Transport Mechanisms

Science.gov (United States)

Dillon, B.; Strom, K.

2017-12-01

Advances in microelectromechanical systems (MEMs) in the past decade have lead to the development of various instrumented or "smart" particles for use in the study of alluvial transport. The goal of many of these devices is to collect data on the interaction between hydrodynamic turbulence and individual sediment particles. Studying this interaction provides a basis to better understand entrainment and deposition processes which leads to better predictive morphologic and transport models. In collecting data on these processes, researchers seek to capture the time history of the forces incident on the particle and the particle's reaction. Many methods have been employed to capture this data - miniaturized pressure traps, accelerometers, gyroscopes, MEMs pressure transducers, and cantilevered load cells. However no system to date has been able to capture the pressure forces incident on the particle and its reaction while remaining mobile and of a size and density comparable to most gravels. Advances in the development, deployment, and use of waterproofed laboratory instrumentation have led our research group to develop such a particle. This particle has been used in both laboratory settings and large-scale fluvial environments (coupled with a field-deployable PIV system) to capture data on turbulent erosion processes. This system advances the practice in several ways: 1) It is, at present, the smallest (⌀ 19mm) instrumented erodible particle reported in the literature. 2) It contains novel developments in pressure sensing technology which allow the inclusion of six pressure ports, a 3-axis accelerometer, and a 1-axis gyroscope - all of which can be recorded simultaneously. 3) It expands the researcher's abilities to gather data on phenomena that, previously, have mandated the use of a laboratory scale model. The use of this system has generated observations of the so-called very large scale motions (VLSMs) in a reach of the Virginia section of the New River. Their

An overview of process instrumentation, protective safety interlocks and alarm system at the JET facilities active gas handling system

International Nuclear Information System (INIS)

Skinner, N.; Brennan, P.; Brown, K.; Gibbons, C.; Jones, G.; Knipe, S.; Manning, C.; Perevezentsev, A.; Stagg, R.; Thomas, R.; Yorkshades, J.

2003-01-01

The Joint European Torus (JET) Facilities Active Gas Handling System (AGHS) comprises ten interconnected processing sub-systems that supply, process and recover tritium from gases used in the JET Machine. Operations require a diverse range of process instrumentation to carry out a multiplicity of monitoring and control tasks and approximately 500 process variables are measured. The different types and application of process instruments are presented with specially adapted or custom-built versions highlighted. Forming part of the Safety Case for tritium operations, a dedicated hardwired interlock and alarm system provides an essential safety function. In the event of failure modes, each hardwired interlock will back-up software interlocks and shutdown areas of plant to a failsafe condition. Design of the interlock and alarm system is outlined and general methodology described. Practical experience gained during plant operations is summarised and the methods employed for routine functional testing of essential instrument systems explained

Cost-effective instrumentation and control upgrades for commercial nuclear power plants using surety principles developed at Sandia National Laboratories

International Nuclear Information System (INIS)

Rochau, G.E.; Dalton, L.J.

1997-01-01

Many nuclear power plants use instrument and control systems based on analog electronics. The state of the art in process control and instrumentation has advanced to use digital electronics and incorporate advanced technology. This technology includes distributed microprocessors, fiber optics, intelligent systems (neural networks), and advanced displays. The technology is used to optimize processes and enhance the man-machine interface while maintaining control and safety of the processes. Nuclear power plant operators have been hesitant to install this technology because of the cost and uncertainty in the regulatory process. This technology can be directly applied in an operating nuclear power plant provided a surety principle-based open-quotes administratorclose quotes hardware system is included in parallel with the upgrade Sandia National Laboratories has developed a rigorous approach to High Consequence System Surety (HCSS). This approach addresses the key issues of safety, security, and control while satisfying requirements for reliability and quality. HCSS principles can be applied to nuclear power plants in a manner that allows the off-the-shelf use of process control instrumentation while maintaining a high level of safety and enhancing the plant performance. We propose that an HCSS administrator be constructed as a standardized approach to address regulatory issues. Such an administrator would allow a plant control system to be constructed with commercially available, state-of-the-art equipment and be customized to the needs of the individual plant operator

Safety design analyses of Korea Advanced Liquid Metal Reactor

International Nuclear Information System (INIS)

Suk, S.D.; Park, C.K.

2000-01-01

The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

Advances in Miniaturized Instruments for Genomics

Directory of Open Access Journals (Sweden)

Cihun-Siyong Alex Gong

2014-01-01

Full Text Available In recent years, a lot of demonstrations of the miniaturized instruments were reported for genomic applications. They provided the advantages of miniaturization, automation, sensitivity, and specificity for the development of point-of-care diagnostics. The aim of this paper is to report on recent developments on miniaturized instruments for genomic applications. Based on the mature development of microfabrication, microfluidic systems have been demonstrated for various genomic detections. Since one of the objectives of miniaturized instruments is for the development of point-of-care device, impedimetric detection is found to be a promising technique for this purpose. An in-depth discussion of the impedimetric circuits and systems will be included to provide total consideration of the miniaturized instruments and their potential application towards real-time portable imaging in the “-omics” era. The current excellent demonstrations suggest a solid foundation for the development of practical and widespread point-of-care genomic diagnostic devices.

Foundational development of an advanced nuclear reactor integrated safety code

International Nuclear Information System (INIS)

Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

2010-01-01

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

Foundational development of an advanced nuclear reactor integrated safety code.

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

2010-02-01

This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

Advances in control and instrumentation

International Nuclear Information System (INIS)

Surendar, Ch.

1994-01-01

Control and instrumentation systems have seen significant changes from pneumatic to electronic with the advent of transistors and integrated circuits. Miniaturization was realised. With the introduction of microprocessors there has been a revolutionary change in the approach in instrumentation and control systems in the areas of sensors, data acquisition/transmission, processing for control, and presentation of the information to the operator. An effort is made to give some insight into these areas, with some idea of the advantages to which these systems are being put to use in the nuclear facilities, particularly nuclear power reactors. (author)

Fiscal 2000 pioneering research report on the research on advanced safety helicopter; 2000 nendo advanced safety helicopter no chosa kenkyu hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

A survey was conducted concerning helicopter operating environments and the trends of related technologies in Japan and abroad, and the needs and seeds were grasped. Research was made to study technical problems and measures to solve them for the development of a safe, low-noise, and low-cost next-generation advanced safety helicopter (ASH). A market research was conducted on traffic systems in the future, state of aviation-related infrastructures and their future, current state of people's daily life which centers about locomotion, and the effect that ASH would impose on society. A technical research was carried out relative to flight safety, which involved EVS (enhanced vision system), information display system for helicopters, collision avoidance advisory for pilots, air collision prevention system/surveillance system for helicopters, obstacle detection/warning system for helicopters, blade deicing system for helicopters, and so forth. Detailed investigations were also conducted for technologies for reduction in the manufacturing, maintenance, and development costs, and for reduction in noise. (NEDO)

Proposal for a advanced PWR core with adequate characteristics for passive safety concept

International Nuclear Information System (INIS)

Perrotta, Jose Augusto

1999-01-01

This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

Astrobiology Sample Analysis Program (ASAP) for Advanced Life Detection Instrumentation Development and Calibration

Science.gov (United States)

Glavin, Daniel; Brinkerhoff, Will; Dworkin, Jason; Eigenbrode, Jennifer; Franz, Heather; Mahaffy, Paul; Stern, Jen; Blake, Daid; Sandford, Scott; Fries, marc;

Instruments for the promotion of environmental innovations. Stocktaking, evaluation and deficit analysis; Instrumente zur Foerderung von Umweltinnovationen. Bestandsaufnahme, Bewertung und Defizitanalyse

Energy Technology Data Exchange (ETDEWEB)

Rennings, Klaus; Rammer, Christian; Oberndorfer, Ulrich [Zentrum fuer Europaeische Wirtschaftsforschung (ZEW), GmbH, Mannheim (DE)] (and others)

2008-03-15

In the ecological report 2006 of the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (Berlin, Federal Republic of Germany) with the title ''Environment - Innovation - Occupation'', the Federal Ministry outlines the concept of an innovative oriented environmental policy. An innovative environmental policy contributes to the reduction of the environmental pollution and the ecological damages of industrial production. In this background, the question arises, how the environmental policy can support appropriate innovative activities. For this, the research project under consideration presents the following contributions: (a) Overview of the most important instruments for the promotion of environmental innovations in Germany; (b) Estimation of the effectiveness of the instruments including gap analysis and deficit analysis; (c) Presentation of foreign Best Practice Examples (instruments and environmental political beginnings); (d) Proposals for priority environmental political starting points to the advancement of the equipment and pointing out the further research need.

Assessment of the wish to hasten death in patients with advanced disease: A systematic review of measurement instruments.

Science.gov (United States)

Bellido-Pérez, Mercedes; Monforte-Royo, Cristina; Tomás-Sábado, Joaquín; Porta-Sales, Josep; Balaguer, Albert

2017-06-01

Patients with advanced conditions may present a wish to hasten death. Assessing this wish is complex due to the nature of the phenomenon and the difficulty of conceptualising it. To identify and analyse existing instruments for assessing the wish to hasten death and to rate their reported psychometric properties. Systematic review based on PRISMA guidelines. The COnsensus-based Standards for the selection of health Measurement INstruments checklist was used to evaluate the methodological quality of validation studies and the measurement properties of the instrument described. The CINAHL, PsycINFO, Pubmed and Web of Science databases were searched from inception to November 2015. A total of 50 articles involving assessment of the wish to hasten death were included. Eight concerned instrument validation and were evaluated using COnsensus-based Standards for the selection of health Measurement INstruments criteria. They reported data for between two and seven measurement properties, with ratings between fair and excellent. Of the seven instruments identified, the Desire for Death Rating Scale or the Schedule of Attitudes toward Hastened Death feature in 48 of the 50 articles. The Schedule of Attitudes toward Hastened Death is the most widely used and is the instrument whose psychometric properties have been most often analysed. Versions of the Schedule of Attitudes toward Hastened Death are available in five languages other than the original English. This systematic review has analysed existing instruments for assessing the wish to hasten death. It has also explored the methodological quality of studies that have examined the measurement properties of these instruments and offers ratings of the reported properties. These results will be useful to clinicians and researchers with an interest in a phenomenon of considerable relevance to advanced patients.

Regulatory perspective on digital instrumentation and control systems for future advanced nuclear power plants

International Nuclear Information System (INIS)

Chiramal, M.

1993-01-01

This paper deals with the question of using digital technology in instrumentation and control systems for modern nuclear power reactors. The general opinion in the industry and among NRC staff is that such technology provides the opportunity for enhanced safety and reliable reactor operations. The major concern is the safe application of this technology so as to avoid common mode or common cause failures in systems. There are great differences between digital and analog system components. SECY-91-292 identifies some general regulatory concerns with regard to digital systems. There is clearly a lack of adequate regulatory direction on the application of digital equipment at this time, but the issue is being addressed by the industry, outside experts, and NRC staff. NRC staff presents a position on the issue of defense-in-depth and diversity with regard to insuring plant safety. Independent manual controls and readouts must be available to allow safe shutdown and monitoring of the plant in the event of safety system failures

Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

International Nuclear Information System (INIS)

Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

1979-01-01

The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

Instrumentation, Control, and Intelligent Systems

Energy Technology Data Exchange (ETDEWEB)

2005-09-01

Abundant and affordable energy is required for U.S. economic stability and national security. Advanced nuclear power plants offer the best near-term potential to generate abundant, affordable, and sustainable electricity and hydrogen without appreciable generation of greenhouse gases. To that end, Idaho National Laboratory (INL) has been charged with leading the revitalization of nuclear power in the U.S. The INL vision is to become the preeminent nuclear energy laboratory with synergistic, world-class, multi-program capabilities and partnerships by 2015. The vision focuses on four essential destinations: (1) Be the preeminent internationally-recognized nuclear energy research, development, and demonstration laboratory; (2) Be a major center for national security technology development and demonstration; (3) Be a multi-program national laboratory with world-class capabilities; (4) Foster academic, industry, government, and international collaborations to produce the needed investment, programs, and expertise. Crucial to that effort is the inclusion of research in advanced instrumentation, control, and intelligent systems (ICIS) for use in current and advanced power and energy security systems to enable increased performance, reliability, security, and safety. For nuclear energy plants, ICIS will extend the lifetime of power plant systems, increase performance and power output, and ensure reliable operation within the system's safety margin; for national security applications, ICIS will enable increased protection of our nation's critical infrastructure. In general, ICIS will cost-effectively increase performance for all energy security systems.

Nuclear power plant control and instrumentation 1982. Proceedings of an international symposium on nuclear power plant control and instrumentation

International Nuclear Information System (INIS)

1983-01-01

Ever increasing demands for nuclear power plant safety and availability imply a need for the introduction of modern measurement and control methods, together with data processing techniques based on the latest advances in electronic components, transducers and computers. Nuclear power plant control and instrumentation is therefore an extremely rapidly developing field. The present symposium, held in Munich, FR Germany, was prepared with the help of the IAEA International Working Group on Nuclear Power Plant Control and Instrumentation and organized in close co-operation with the Gesellschaft fur Reaktorsicherheit, Federal Republic of Germany. A number of developments were highlighted at the Munich symposium: - The increased use of computers can bring clear advantages and this technique is now proven as a tool for supervising and controlling plant operation. Advanced computerized systems for operator support are being developed on a large scale in many countries. The progress in this field is quite obvious, especially in disturbance analysis, safety parameter display, plant operator guidance and plant diagnostics. The new trend of introducing computers and microprocessors in protection systems makes it easy to implement 'defence-in-depth' strategies which give better assurance of correct system responses and also prevent unnecessary reactor trips, thus improving plant availability. The introduction of computerized systems for control of reactor power, reactor water level and reactor pressure as well as for reactor start-up and shut-down could improve the reliability and availability of nuclear power plants. The rapid technical development in the area of control and instrumentation makes it necessary to plan for at least one replacement of obsolete equipment in the course of the 30 years lifetime of a nuclear power plant and retrofitting of currently operating reactors with new control systems. Major design improvements and regulatory requirements also require

Advances in High-Throughput Speed, Low-Latency Communication for Embedded Instrumentation (7th Annual SFAF Meeting, 2012)

Energy Technology Data Exchange (ETDEWEB)

Jordan, Scott

2012-06-01

Scott Jordan on "Advances in high-throughput speed, low-latency communication for embedded instrumentation" at the 2012 Sequencing, Finishing, Analysis in the Future Meeting held June 5-7, 2012 in Santa Fe, New Mexico.

Nanosurveyor 2: A Compact Instrument for Nano-Tomography at the Advanced Light Source

Science.gov (United States)

Celestre, Richard; Nowrouzi, Kasra; Shapiro, David A.; Denes, Peter; Joseph, John M.; Schmid, Andreas; Padmore, Howard A.

2017-06-01

The Advanced Light Source has developed a compact tomographic microscope based on soft x-ray ptychography for the study of nanoscale materials [1,2]. The microscope utilizes the sample manipulator mechanism from a commercial TEM coupled with laser interferometric feedback for zone plate positioning and a fast frame rate charge-coupled device detector for soft x-ray diffraction measurements. The microscope has achieved point to point (25 nm steps) scan rates of greater than 120 Hz with a positioning accuracy of better than 1 nm RMS. The instrument will enable the use of commercially available sample holders compatible with FEI transmission electron microscopes thus also allowing in-situ measurement of samples using both soft x-rays and electrons. This instrument is a refinement of a currently commissioned instrument called The Nanosurveyor, which has demonstrated resolution of better than 10 nm in two dimensions using 750 eV x-rays. Once moved to the new Coherent Scattering and Microscopy beamline it will enable spectromicroscopy and tomography of nano-materials with wavelength limited spatial resolution.

Conceptual safety design analysis of Korea advanced liquid metal reactor

International Nuclear Information System (INIS)

Suk, S. D.; Park, C. K.

1999-01-01

The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

Advanced Photon Source experimental beamline Safety Assessment Document: Addendum to the Advanced Photon Source Accelerator Systems Safety Assessment Document (APS-3.2.2.1.0)

International Nuclear Information System (INIS)

1995-01-01

This Safety Assessment Document (SAD) addresses commissioning and operation of the experimental beamlines at the Advanced Photon Source (APS). Purpose of this document is to identify and describe the hazards associated with commissioning and operation of these beamlines and to document the measures taken to minimize these hazards and mitigate the hazard consequences. The potential hazards associated with the commissioning and operation of the APS facility have been identified and analyzed. Physical and administrative controls mitigate identified hazards. No hazard exists in this facility that has not been previously encountered and successfully mitigated in other accelerator and synchrotron radiation research facilities. This document is an updated version of the APS Preliminary Safety Analysis Report (PSAR). During the review of the PSAR in February 1990, the APS was determined to be a Low Hazard Facility. On June 14, 1993, the Acting Director of the Office of Energy Research endorsed the designation of the APS as a Low Hazard Facility, and this Safety Assessment Document supports that designation

Comparison of a Traditional Probabilistic Risk Assessment Approach with Advanced Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis L; Mandelli, Diego; Zhegang Ma

2014-11-01

As part of the Light Water Sustainability Program (LWRS) [1], the purpose of the Risk Informed Safety Margin Characterization (RISMC) [2] Pathway research and development (R&D) is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario. We also describe our approach we are using to represent advanced flooding analysis.

Outline of the requirements of application of computer based instrumentation and control systems in the systems important to safety on Bohunice NPPs

International Nuclear Information System (INIS)

Bacurik, J.

1997-01-01

The most important regulatory requirements and issues are described related to the review, evaluation and assessment of computer-based safety-related IandC systems, with emphasis on safety instrumentation and control. These aspects include safety classification and categorization of IandC, ranking of applicable codes and standards, design evaluation on the system level, and software assessment. (author)

Motorcycle That See: Multifocal Stereo Vision Sensor for Advanced Safety Systems in Tilting Vehicles

Directory of Open Access Journals (Sweden)

Gustavo Gil

2018-01-01

Full Text Available Advanced driver assistance systems, ADAS, have shown the possibility to anticipate crash accidents and effectively assist road users in critical traffic situations. This is not the case for motorcyclists, in fact ADAS for motorcycles are still barely developed. Our aim was to study a camera-based sensor for the application of preventive safety in tilting vehicles. We identified two road conflict situations for which automotive remote sensors installed in a tilting vehicle are likely to fail in the identification of critical obstacles. Accordingly, we set two experiments conducted in real traffic conditions to test our stereo vision sensor. Our promising results support the application of this type of sensors for advanced motorcycle safety applications.

Motorcycles that See: Multifocal Stereo Vision Sensor for Advanced Safety Systems in Tilting Vehicles

Science.gov (United States)

2018-01-01

Advanced driver assistance systems, ADAS, have shown the possibility to anticipate crash accidents and effectively assist road users in critical traffic situations. This is not the case for motorcyclists, in fact ADAS for motorcycles are still barely developed. Our aim was to study a camera-based sensor for the application of preventive safety in tilting vehicles. We identified two road conflict situations for which automotive remote sensors installed in a tilting vehicle are likely to fail in the identification of critical obstacles. Accordingly, we set two experiments conducted in real traffic conditions to test our stereo vision sensor. Our promising results support the application of this type of sensors for advanced motorcycle safety applications. PMID:29351267

Motorcycle That See: Multifocal Stereo Vision Sensor for Advanced Safety Systems in Tilting Vehicles.

Science.gov (United States)

Gil, Gustavo; Savino, Giovanni; Piantini, Simone; Pierini, Marco

2018-01-19

Advanced driver assistance systems, ADAS, have shown the possibility to anticipate crash accidents and effectively assist road users in critical traffic situations. This is not the case for motorcyclists, in fact ADAS for motorcycles are still barely developed. Our aim was to study a camera-based sensor for the application of preventive safety in tilting vehicles. We identified two road conflict situations for which automotive remote sensors installed in a tilting vehicle are likely to fail in the identification of critical obstacles. Accordingly, we set two experiments conducted in real traffic conditions to test our stereo vision sensor. Our promising results support the application of this type of sensors for advanced motorcycle safety applications.

Advances in instrumentation for nuclear astrophysics

Directory of Open Access Journals (Sweden)

S. D. Pain

2014-04-01

Full Text Available The study of the nuclear physics properties which govern energy generation and nucleosynthesis in the astrophysical phenomena we observe in the universe is crucial to understanding how these objects behave and how the chemical history of the universe evolved to its present state. The low cross sections and short nuclear lifetimes involved in many of these reactions make their experimental determination challenging, requiring developments in beams and instrumentation. A selection of developments in nuclear astrophysics instrumentation is discussed, using as examples projects involving the nuclear astrophysics group at Oak Ridge National Laboratory. These developments will be key to the instrumentation necessary to fully exploit nuclear astrophysics opportunities at the Facility for Rare Isotope Beams which is currently under construction.

Advanced Neutron Source (ANS) Project progress report

International Nuclear Information System (INIS)

McBee, M.R.; Chance, C.M.

1990-04-01

This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I ampersand C research and development; facility concepts; design; and safety

Uranium Mill Tailings Remedial Action Project Safety Advancement Field Effort (SAFE) Program

Energy Technology Data Exchange (ETDEWEB)

1994-02-01

In 1992, the Uranium Mill Tailings Remedial Action (UMTRA) Project experienced several health and safety related incidents at active remediation project sites. As a result, the U.S. Department of Energy (DOE) directed the Technical Assistance Contractor (TAC) to establish a program increasing the DOE`s overall presence at operational remediation sites to identify and minimize risks in operations to the fullest extent possible (Attachments A and B). In response, the TAC, in cooperation with the DOE and the Remedial Action Contractor (RAC), developed the Safety Advancement Field Effort (SAFE) Program.

Uranium Mill Tailings Remedial Action Project Safety Advancement Field Effort (SAFE) Program

International Nuclear Information System (INIS)

1994-02-01

In 1992, the Uranium Mill Tailings Remedial Action (UMTRA) Project experienced several health and safety related incidents at active remediation project sites. As a result, the U.S. Department of Energy (DOE) directed the Technical Assistance Contractor (TAC) to establish a program increasing the DOE's overall presence at operational remediation sites to identify and minimize risks in operations to the fullest extent possible (Attachments A and B). In response, the TAC, in cooperation with the DOE and the Remedial Action Contractor (RAC), developed the Safety Advancement Field Effort (SAFE) Program

Monograph on safety in high power and high energy advanced technologies and medical applications of lasers

International Nuclear Information System (INIS)

2016-01-01

This monograph is intended for creating awareness amongst the safety and health professionals of nuclear and radiation facilities on hazards involved in high power and high energy advanced technologies as well as on how development of advanced technologies can benefit the common people

Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

International Nuclear Information System (INIS)

Hadid Subki, M.; )

2014-01-01

The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

Upgrading instrumentation and control systems for plant safety and operation

International Nuclear Information System (INIS)

Martin, M.; Prehler, H.J.; Schramm, W.

1997-01-01

Upgrading the electrical systems and instrumentation and control systems has become increasingly more important in the past few years for nuclear power plants currently in operation. As the requirements to be met in terms of plant safety and availability have become more stringent in the past few years, Western plants built in the sixties and seventies have been the subject of manifold backfitting and upgrading measures in the past. In the meantime, however, various nuclear power plants are facing much more thorough upgrading phases because of the difficulties in obtaining spare parts for older equipment systems. As digital technology has become widespread in many areas because of its advantages, and as applications are continuously expanding, conventional equipment and systems are losing more and more ground as a consequence of decreasing demand. Merely because of the pronounced decline in demand for conventional electronic components it is possible for equipment manufacturers to guarantee spare parts deliveries for older systems only for specific future periods of time. In addition, one-off manufacture entails high costs in purchases of spare parts. As a consequence of current thinking more and more focusing on availability and economy, upgrading of electrical systems and instrumentation and control systems is becoming a more and more topical question, for older plants even to ensure completion of full service life. (orig.) [de

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs.

International legal instruments promoting synergy's in nuclear safety, security and safeguards: myth of reality?

International Nuclear Information System (INIS)

Vasmant, A.

2009-01-01

The purpose of this article is to assess the existing synergies between nuclear safety, nuclear security and non-proliferation/safeguards resulting from the adoption of international legal instruments. Keeping in mind that a synergy is the extra success achieved by two or more elements of a system working together instead of on their own, this paper will try to evaluate the possibility of a so-called '3 S' approach to optimize the benefits so defined. to achieve this, Part 1 focuses on the history of the three regimes and their major features, while Part 2, 3 and 4 explore the various benefits of, limits to, synergies between the nuclear safety, nuclear security and safeguards regimes. Part 5 describes the potential '3 S' approach in international nuclear law. (N.C.)

Advancing the Science of Developmental Neurotoxicity (DNT) Testing for Better Safety Evaluation

DEFF Research Database (Denmark)

Bal-Price, Anna; Coecke, Sandra; Costa, Lucio

2012-01-01

Bal-Price AK, Coecke S, Costa L, Crofton KM, Fritsche E, Goldberg A, Grandjean P, Lein PJ, Li A, Lucchini R, Mundy WR, Padilla S, Persico A, Seiler AEM, Kreysa J. Conference Report: Advancing the Science of Developmental Neurotoxicity (DNT) Testing for Better Safety Evaluation. Altex 2012: 29: 202-15....

A survey on the development of advanced instrumentation and control system in NPP

International Nuclear Information System (INIS)

Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan

1993-12-01

Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: 1) N4 of EDF in France, 2) AP 600 of Westinghouse in USA, 3) NUPLEX-80+ of ABB-CE in USA, 4) CANDU in Canada, 5) Ohi 3 and 4, APWR and ABWR in Japan, 6) Belt-D in Germany, 7) Sizewell B in Britain, 8) Halden Reactor Projector in Norway, 9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author)

A survey on the development of advanced instrumentation and control system in NPP

Energy Technology Data Exchange (ETDEWEB)

Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-12-01

Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: (1) N4 of EDF in France, (2) AP 600 of Westinghouse in USA, (3) NUPLEX-80+ of ABB-CE in USA, (4) CANDU in Canada, (5) Ohi 3 and 4, APWR and ABWR in Japan, (6) Belt-D in Germany, (7) Sizewell B in Britain, (8) Halden Reactor Projector in Norway, (9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author).

Experience and regulatory activities on advanced instrumentation and control systems applied to nuclear power plants in Korea

International Nuclear Information System (INIS)

Kim, B.R.

1995-01-01

This paper describes the status for applying microprocessor-based systems to nuclear power plants in Korea and the regulatory activities performed by Korea Institute of Nuclear Safety (KINS). And this presents the development of safety and regulatory technology for advanced I and C systems that has been carried out as a part of the next generation reactor development program in Korea. (author). 3 refs, 4 figs, 1 tab

Atmospheric effects on laser eye safety and damage to instrumentation

Science.gov (United States)

Zilberman, Arkadi; Kopeika, Natan S.

2017-10-01

Electro-optical sensors as well as unprotected human eyes are extremely sensitive to laser radiation and can be permanently damaged from direct or reflected beams. Laser detector/eye hazard depends on the interaction between the laser beam and the media in which it traverses. The environmental conditions including terrain features, atmospheric particulate and water content, and turbulence, may alter the laser's effect on the detector/eye. It is possible to estimate the performance of an electro-optical system as long as the atmospheric propagation of the laser beam can be adequately modeled. More recent experiments and modeling of atmospheric optics phenomena such as inner scale effect, aperture averaging, atmospheric attenuation in NIR-SWIR, and Cn2 modeling justify an update of previous eye/detector safety modeling. In the present work, the influence of the atmospheric channel on laser safety for personnel and instrumentation is shown on the basis of theoretical and experimental data of laser irradiance statistics for different atmospheric conditions. A method for evaluating the probability of damage and hazard distances associated with the use of laser systems in a turbulent atmosphere operating in the visible and NIR-SWIR portions of the electromagnetic spectrum is presented. It can be used as a performance prediction model for directed energy engagement of ground-based or air-based systems.

Earth Observing System (EOS) Advanced Microwave Sounding Unit-A (AMSU-A): Instrumentation interface control document

Science.gov (United States)

1994-01-01

This Interface Control Document (ICD) defines the specific details of the complete accomodation information between the Earth Observing System (EOS) PM Spacecraft and the Advanced Microwave Sounding Unit (AMSU-A)Instrument. This is the first submittal of the ICN: it will be updated periodically throughout the life of the program. The next update is planned prior to Critical Design Review (CDR).

Safeguarding the functions and performance of instrumentation and control systems

International Nuclear Information System (INIS)

Koehler, M.; Schoerner, O.

1996-01-01

Based on an analysis of the existing nuclear power plant control technology, the necessity of providing in the medium-term advanced and future-oriented, digital control system, both for normal operation and for safety-relevant tasks of the reactor and safety control systems. Siemens KWU has been promoting the development, review and marketing of the digital instrumentation and control systems called TELEPERM XS and TELEPERM XP in addition to the measures taken for safeguarding the functions of existing, wired systems. The paper briefly explains the performance and advantages of digital systems and the progress in approval and pioneering of the TELEPERM XS safety control system. Many examples discussed show the diversity of applications of the systems both in new reactor plants and as retrofitting measures, for KWU power plants and those of other manufacturers. (orig.) [de

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

International Nuclear Information System (INIS)

Hallbert, Bruce P.; Persensky, J.J.; Smidts, Carol; Aldemir, Tunc; Naser, Joseph

2009-01-01

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R and D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I and C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I and C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R and D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R and D to develop scientific knowledge in the areas of: (1) Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs); (2) Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information; (3) New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

Energy Technology Data Exchange (ETDEWEB)

Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

2009-08-01

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation

Advances in constant-velocity Moessbauer instrumentation

International Nuclear Information System (INIS)

Veiga, A.; Martinez, N.; Zelis, P. Mendoza; Pasquevich, G. A.; Sanchez, F. H.

2006-01-01

A prototype of a programmable constant-velocity scaler is presented. This instrument allows the acquisition of partial Moessbauer spectra in selected energy regions using standard drivers and transducers. It can be fully operated by a remote application, thus data acquisition can be automated. The instrument consists of a programmable counter and a constant-velocity reference. The reference waveform generator is amplitude modulated with 13-bit resolution, and is programmable in a wide range of frequencies and waveforms in order to optimize the performance of the transducer. The counter is compatible with most standard SCA, and is configured as a rate-meter that provides counts per selectable time slice at the programmed velocity. As a demonstration of the instrument applications, a partial Moessbauer spectrum of a natural iron foil was taken. Only positive energies were studied in 512 channels, accumulating 20 s per channel. A line width of 0.20 mm/s was achieved, performing with an efficiency of 80%.

Advanced Neutron Source (ANS) Project Progress report, FY 1991

International Nuclear Information System (INIS)

Campbell, J.H.; Selby, D.L.; Harrington, R.M.; Thompson, P.B.

1992-01-01

This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I ampersand C Research and Development; Design; and Safety

EPRI's nuclear power plant instrumentation and control program and its applicability to advanced reactors

International Nuclear Information System (INIS)

Naser, J.; Torok, R.; Wilkinson, D.

1997-01-01

I ampersand C systems in nuclear power plants need to be upgraded over the lifetime of the plant in a reliable and cost-effective manner to replace obsolete equipment, to reduce O ampersand M costs, to improve plant performance, and to maintain safety. This applies to operating plants now and will apply to advanced reactors in the future. The major drivers for the replacement of the safety, control, and information systems in nuclear power plants are the obsolescence of the existing hardware and the need for more cost-effective power production. Competition between power producers is dictating more cost-effective power production. The increasing O ampersand M costs to maintain systems experiencing obsolescence problems is counter to the needs for more cost-effective power production and improved competitiveness. This need for increased productivity applies to government facilities as well as commercial plants. Increasing competition will continue to be a major factor in the operation of both operating plants and advanced reactors. It will continue to dictate the need for improved productivity and cost-effectiveness. EPRI and its member nuclear utilities are working together on an industry wide I ampersand C Program to address I ampersand C issues and to develop cost-effective solutions. A majority of the I ampersand C products and demonstrations being developed under this program will benefit advanced reactors in both the design and operational phases of their life cycle as well as it will benefit existing plants. 20 refs

Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

International Nuclear Information System (INIS)

Hallbert, Bruce P.; Kenneth, Thomas

2014-01-01

The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security

Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Hallbert, Bruce P.; Kenneth, Thomas [Idaho National Laboratory, Idaho (United States)

2014-08-15

The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

Power station instrumentation

International Nuclear Information System (INIS)

Jervis, M.W.

1993-01-01

Power stations are characterized by a wide variety of mechanical and electrical plant operating with structures, liquids and gases working at high pressures and temperatures and with large mass flows. The voltages and currents are also the highest that occur in most industries. In order to achieve maximum economy, the plant is operated with relatively small margins from conditions that can cause rapid plant damage, safety implications, and very high financial penalties. In common with other process industries, power stations depend heavily on control and instrumentation. These systems have become particularly significant, in the cost-conscious privatized environment, for providing the means to implement the automation implicit in maintaining safety standards, improving generation efficiency and reducing operating manpower costs. This book is for professional instrumentation engineers who need to known about their use in power stations and power station engineers requiring information about the principles and choice of instrumentation available. There are 8 chapters; chapter 4 on instrumentation for nuclear steam supply systems is indexed separately. (Author)

IAEA safety standards and approach to safety of advanced reactors

International Nuclear Information System (INIS)

Gasparini, M.

2004-01-01

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

Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

International Nuclear Information System (INIS)

G. L. Sharp; R. T. McCracken

2004-01-01

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

Advanced safety management systems for maintenance of pipeline integrity

International Nuclear Information System (INIS)

Borysiewicz, M.; Potempski, S.

2005-01-01

One of the duties of the pipeline's operator is to introduce means for protection of human safety and the environment. This should be reflected in preparation of comprehensive Risk Management System with its key element Activity Programme for Management of Pipeline Integrity. In the paper such programme has been described taking into account law regulations and practical activities undertaken in technologically advanced countries (mainly USA and EU), where such solutions are implemented in routine operations. Possible solutions of realization of all elements of the programme, as well as information on utilization of computer aided support have been also included. (authors)

Advanced CFD simulation for the assessment of nuclear safety issues at EDF. Some examples

International Nuclear Information System (INIS)

Vare, Christophe

2014-01-01

EDF R and D has computer power that puts it amongst the top industrial research centers in the world. Its supercomputers and in-house codes as well as its experts represent important capabilities to support EDF activities (safety analyses, support to the design of new reactors, analysis of accidental situations non reproducible by experiments, better understanding of physics or complex system response, effects of uncertainties and identification of prominent parameters, qualification and optimization of processes and materials...). Advanced numerical simulation is a powerful tool allowing EDF to increase its competitiveness, improve its performance and the safety of its plants. On this issue, EDF made the choice to develop its own in-house codes, instead of using commercial software, in order to be able to capitalize its expertise and methodologies. This choice allowed as well easier technological transfer to the concerned business units or engineering divisions, fast adaptation of our simulation tools to emerging needs and the development of specific physics or functionalities not addressed by the commercial offer. During the last ten years, EDF has decided to open its in-house codes, through the Open Source way. This is the case for Code – Aster (structure analysis), Code – Saturne (computational fluid dynamics, CFD), TELEMAC (flow calculations in aquatic environment), SALOME (generic platform for Pre and Post-Processing) and SYRTHES (heat transfer in complex geometries), among others. The 3 open source software: Code – Aster, Code – Saturne and TELEMAC, are certified by the French Nuclear Regulatory Authority for many «Important to Safety» studies. Advanced simulation, which treats complex, multi-field and multi-physics problems, is of great importance for the assessment of nuclear safety issues. This paper will present 2 examples of advanced simulation using Code – Saturne for safety issues of nuclear power plants in the fields of R and D and

Using level-I PRA for enhanced safety of the advanced neutron source reactor

International Nuclear Information System (INIS)

Ramsey, C.T.; Linn, M.A.

1995-01-01

The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

Quality assurance for software important to safety

International Nuclear Information System (INIS)

2000-01-01

Software applications play an increasingly relevant role in nuclear power plant systems. This is particularly true of software important to safety used in both: calculations for the design, testing and analysis of nuclear reactor systems (design, engineering and analysis software); and monitoring, control and safety functions as an integral part of the reactor systems (monitoring, control and safety system software). Computer technology is advancing at a fast pace, offering new possibilities in nuclear reactor design, construction, commissioning, operation, maintenance and decommissioning. These advances also present new issues which must be considered both by the utility and by the regulatory organization. Refurbishment of ageing instrumentation and control systems in nuclear power plants and new safety related application areas have emerged, with direct (e.g. interfaces with safety systems) and indirect (e.g. operator intervention) implications for safety. Currently, there exist several international standards and guides on quality assurance for software important to safety. However, none of the existing documents provides comprehensive guidance to the developer, manager and regulator during all phases of the software life-cycle. The present publication was developed taking into account the large amount of available documentation, the rapid development of software systems and the need for updated guidance on h ow to do it . It provides information and guidance for defining and implementing quality assurance programmes covering the entire life-cycle of software important to safety. Expected users are managers, performers and assessors from nuclear utilities, regulatory bodies, suppliers and technical support organizations involved with the development and use of software applied in nuclear power plants

ECLSS Integration Analysis: Advanced ECLSS Subsystem and Instrumentation Technology Study for the Space Exploration Initiative

Science.gov (United States)

1990-01-01

In his July 1989 space policy speech, President Bush proposed a long range continuing commitment to space exploration and development. Included in his goals were the establishment of permanent lunar and Mars habitats and the development of extended duration space transportation. In both cases, a major issue is the availability of qualified sensor technologies for use in real-time monitoring and control of integrated physical/chemical/biological (p/c/b) Environmental Control and Life Support Systems (ECLSS). The purpose of this study is to determine the most promising instrumentation technologies for future ECLSS applications. The study approach is as follows: 1. Precursor ECLSS Subsystem Technology Trade Study - A database of existing and advanced Atmosphere Revitalization (AR) and Water Recovery and Management (WRM) ECLSS subsystem technologies was created. A trade study was performed to recommend AR and WRM subsystem technologies for future lunar and Mars mission scenarios. The purpose of this trade study was to begin defining future ECLSS instrumentation requirements as a precursor to determining the instrumentation technologies that will be applicable to future ECLS systems. 2. Instrumentation Survey - An instrumentation database of Chemical, Microbial, Conductivity, Humidity, Flowrate, Pressure, and Temperature sensors was created. Each page of the sensor database report contains information for one type of sensor, including a description of the operating principles, specifications, and the reference(s) from which the information was obtained. This section includes a cursory look at the history of instrumentation on U.S. spacecraft. 3. Results and Recommendations - Instrumentation technologies were recommended for further research and optimization based on a consideration of both of the above sections. A sensor or monitor technology was recommended based on its applicability to future ECLS systems, as defined by the ECLSS Trade Study (1), and on whether its

RISMC advanced safety analysis project plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

International Nuclear Information System (INIS)

Szilard, Ronaldo H; Smith, Curtis L; Youngblood, Robert

2014-01-01

In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (@@@why is this important?@@@) that will make the case for stakeholder's use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable @@use case@@@ demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

The ConCom Safety Management Scale: developing and testing a measurement instrument for control-based and commitment-based safety management approaches in hospitals.

Science.gov (United States)

Alingh, Carien W; Strating, Mathilde M H; van Wijngaarden, Jeroen D H; Paauwe, Jaap; Huijsman, Robbert

2018-03-06

Nursing management is considered important for patient safety. Prior research has predominantly focused on charismatic leadership styles, although it is questionable whether these best characterise the role of nurse managers. Managerial control is also relevant. Therefore, we aimed to develop and test a measurement instrument for control-based and commitment-based safety management of nurse managers in clinical hospital departments. A cross-sectional survey design was used to test the newly developed questionnaire in a sample of 2378 nurses working in clinical departments. The nurses were asked about their perceptions of the leadership behaviour and management practices of their direct supervisors. Psychometric properties were evaluated using confirmatory factor analysis and reliability estimates. The final 33-item questionnaire showed acceptable goodness-of-fit indices and internal consistency (Cronbach's α of the subscales range: 0.59-0.90). The factor structure revealed three subdimensions for control-based safety management: (1) stressing the importance of safety rules and regulations; (2) monitoring compliance; and (3) providing employees with feedback. Commitment-based management consisted of four subdimensions: (1) showing role modelling behaviour; (2) creating safety awareness; (3) showing safety commitment; and (4) encouraging participation. Construct validity of the scale was supported by high factor loadings and provided preliminary evidence that control-based and commitment-based safety management are two distinct yet related constructs. The findings were reconfirmed in a cross-validation procedure. The results provide initial support for the construct validity and reliability of our ConCom Safety Management Scale. Both management approaches were found to be relevant for managing patient safety in clinical hospital departments. The scale can be used to deepen our understanding of the influence of patient safety management on healthcare professionals

Balanced Design of Safety Systems of CAREM Advanced Reactor

International Nuclear Information System (INIS)

Grinblat, Pablo; Gimenez, Marcelo; Schlamp, Miguel

2003-01-01

Nuclear Power Plants must meet the performance that the market and the population demand in order to be part of the electricity supply industry.It is related mainly with the results of reactor's economy and safety.New advances in the methodology developed for reactor economic optimization analyzing its safety at an early engineering stage, aiming at balancing these important features of the design, are presented in this work.In particular, the coupling that appears when dimensioning the Emergency Injection System, the Residual Heat Removal System and the containment height of CAREM reactor is described.The new models appended to the computer code that embodies the methodology to balance de designs are shown.Finally the results obtained with the optimizations when applying it are presented.Furthermore, a criterion to establish the maximal diameter for acceptable breaks in RPV's penetrations arises from this work.The application of the methodology and the computer code developed turns out to prove the advantages they provide to reactor design so that the plants are properly balanced and optimized

Safety features and research needs of westinghouse advanced reactors

International Nuclear Information System (INIS)

Carelli, M.D.; Winters, J.W.; Cummins, W.E.; Bruschi, H.J.

2002-01-01

The three Westinghouse advanced reactors - AP600, AP1000 and IRIS - are at different levels of readiness. AP600 has received a Design Certification, its larger size version AP1000 is currently in the design certification process and IRIS has just completed its conceptual design and will initiate soon a licensing pre-application. The safety features of the passive designs AP600/AP1000 are presented, followed by the features of the more revolutionary IRIS, a small size modular integral reactor. A discussion of the IRIS safety by design approach is given. The AP600/AP1000 design certification is backed by completed testing and development which is summarized, together with a research program currently in progress which will extend AP600 severe accident test data to AP1000 conditions. While IRIS will of course rely on applicable AP600/1000 data, a very extensive testing campaign is being planned to address all the unique aspects of its design. Finally, IRIS plans to use a risk-informed approach in its licensing process. (authors)

Advancement of Tools Supporting Improvement of Work Safety in Selected Industrial Company

Science.gov (United States)

Gembalska-Kwiecień, Anna

2018-03-01

In the presented article, the advancement of tools to improve the safety of work in the researched industrial company was taken into consideration. Attention was paid to the skillful analysis of the working environment, which includes the available technologies, work organization and human capital. These factors determine the development of the best prevention activities to minimize the number of accidents.

Research on the evaluation model of the software reliability in nuclear safety class digital instrumentation and control system

International Nuclear Information System (INIS)

Liu Ying; Yang Ming; Li Fengjun; Ma Zhanguo; Zeng Hai

2014-01-01

In order to analyze the software reliability (SR) in nuclear safety class digital instrumentation and control system (D-I and C), firstly, the international software design standards were analyzed, the standards' framework was built, and we found that the D-I and C software standards should follow the NUREG-0800 BTP7-14, according to the NRC NUREG-0800 review of requirements. Secondly, the quantitative evaluation model of SR using Bayesian Belief Network and thirteen sub-model frameworks were established. Thirdly, each sub-models and the weight of corresponding indexes in the evaluation model were analyzed. Finally, the safety case was introduced. The models lay a foundation for review and quantitative evaluation on the SR in nuclear safety class D-I and C. (authors)

Reconstruction of instrumentation and control system (SKR)

International Nuclear Information System (INIS)

Wiening, K.-H.

2001-01-01

For the first time extensive upgrades have been performed in all safety related areas of units with WWER 440/230 reactors. One of the most important actions was the replacement of the safety and safety related instrumentation and control. The state of the art digital safety instrumentation and control system TELEPERM XS has been implemented in units 1 and 2 of the Bohunice V1 power plant. The requirements as deduced from safety assessments conducted by commissions of international experts have been fulfilled, so that Bohunice V1 after this gradual reconstruction has been upgraded to an internationally accepted safety level for the remainder of its service life. (author)

Seismic PSA implementation standards by AESJ and the utilization of the advanced safety examination guideline for seismic design for nuclear power plant

International Nuclear Information System (INIS)

Ebisawa, Katsumi; Hibino, Kenta

2008-01-01

The Advanced Safety Examination Guideline for Seismic Design for Nuclear Power Plant (the advanced safety examination guideline) was worked out on September 19, 2006. In this paper, a summary of the method of probability theory in the advanced safety examination guideline and the Seismic PSA Implementation Standards is stated. On utilization of the probability theory for the advanced safety examination guideline, the uncertainty resulting from the process of the decision of the basic design earthquake ground motion (Ss) is stated to be considered using the proper method. The references of the extra probability for evaluation of earthquake hazard and combination of the working load and the earthquake load are stated. Definition, evaluation method and effort to lower the 'residual risks', and relation between the residual risks and the extra probability of Ss are described. A summary of the earthquake-resistant design for nuclear power facilities is explained by the old guideline. (S.Y.)

EDITORIAL: Advanced Sensors and Instrumentation Systems for the Food and Beverage Industries

Science.gov (United States)

Yan, Yong

2006-02-01

Advanced sensors and instrumentation systems are becoming increasingly important in the classification, characterization, authentication, quality control and safety management of food products and beverages. To bring together industrialists and academic researchers to discuss the latest developments and trends in this particular area, the ISAT (Instrument Science and Technology) Group of the Institute of Physics organized a highly focused one-day technical meeting, which was held at the Rutherford Conference Centre at the Institute of Physics in London on 15 December 2004. The event was co-sponsored by the Measurement, Sensors, Instrumentation and NDT Professional Network of the Institution of Electrical Engineers and the Measurement Science and Technology Panel of the Institute of Measurement and Control. The special feature in this issue (on pages 229 287) brings together a collection of some of the papers that were presented at the event. Also included in the special feature are two relevant papers that were submitted through the usual route. Technical topics covered, though wide ranging as reflected in part by the diversity of the papers, demonstrate recent developments and possible approaches that may offer solutions to a broad range of sensing and measurement problems in the food and beverage industries. The first paper, reported by Sheridan et al, is concerned with the quality monitoring of chicken, sausages and pastry products during their cooking processes using an optical fibre-based sensing system. Carter et al describe how digital imaging and image processing techniques have been applied to achieve the classification and authentication of rice grains. The challenges in the measurement and control of final moisture content in baked food products such as bread and biscuits are addressed and discussed by McFarlane. Juodeikiene et al report their progress in the development of acoustic echolocation-based techniques for the evaluation of porosity and

Advances in safety related maintenance

International Nuclear Information System (INIS)

2000-03-01

The maintenance of systems, structures and components in nuclear power plants (NPPs) plays an important role in assuring their safe and reliable operation. Worldwide, NPP maintenance managers are seeking to reduce overall maintenance costs while maintaining or improving the levels of safety and reliability. Thus, the issue of NPP maintenance is one of the most challenging aspects of nuclear power generation. There is a direct relation between safety and maintenance. While maintenance alone (apart from modifications) will not make a plant safer than its original design, deficient maintenance may result in either an increased number of transients and challenges to safety systems or reduced reliability and availability of safety systems. The confidence that NPP structures, systems and components will function as designed is ultimately based on programmes which monitor both their reliability and availability to perform their intended safety function. Because of this, approaches to monitor the effectiveness of maintenance are also necessary. An effective maintenance programme ensures that there is a balance between the improvement in component reliability to be achieved and the loss of component function due to maintenance downtime. This implies that the safety level of an NPP should not be adversely affected by maintenance performed during operation. The nuclear industry widely acknowledges the importance of maintenance in NPP safety and operation and therefore devotes great efforts to develop techniques, methods and tools to aid in maintenance planning, follow-up and optimization, and in assuring the effectiveness of maintenance

Labour analgesia: Recent advances

Science.gov (United States)

Pandya, Sunil T

2010-01-01

Advances in the field of labour analgesia have tread a long journey from the days of ether and chloroform in 1847 to the present day practice of comprehensive programme of labour pain management using evidence-based medicine. Newer advances include introduction of newer techniques like combined spinal epidurals, low-dose epidurals facilitating ambulation, pharmacological advances like introduction of remifentanil for patient-controlled intravenous analgesia, introduction of newer local anaesthetics and adjuvants like ropivacaine, levobupivacaine, sufentanil, clonidine and neostigmine, use of inhalational agents like sevoflourane for patient-controlled inhalational analgesia using special vaporizers, all have revolutionized the practice of pain management in labouring parturients. Technological advances like use of ultrasound to localize epidural space in difficult cases minimizes failed epidurals and introduction of novel drug delivery modalities like patient-controlled epidural analgesia (PCEA) pumps and computer-integrated drug delivery pumps have improved the overall maternal satisfaction rate and have enabled us to customize a suitable analgesic regimen for each parturient. Recent randomized controlled trials and Cochrane studies have concluded that the association of epidurals with increased caesarean section and long-term backache remains only a myth. Studies have also shown that the newer, low-dose regimes do not have a statistically significant impact on the duration of labour and breast feeding and also that these reduce the instrumental delivery rates thus improving maternal and foetal safety. Advances in medical technology like use of ultrasound for localizing epidural space have helped the clinicians to minimize the failure rates, and many novel drug delivery modalities like PCEA and computer-integrated PCEA have contributed to the overall maternal satisfaction and safety. PMID:21189877

Pivotal ERIVANCE basal cell carcinoma (BCC) study: 12-month update of efficacy and safety of vismodegib in advanced BCC.

Science.gov (United States)

Sekulic, Aleksandar; Migden, Michael R; Lewis, Karl; Hainsworth, John D; Solomon, James A; Yoo, Simon; Arron, Sarah T; Friedlander, Philip A; Marmur, Ellen; Rudin, Charles M; Chang, Anne Lynn S; Dirix, Luc; Hou, Jeannie; Yue, Huibin; Hauschild, Axel

2015-06-01

Primary analysis from the pivotal ERIVANCE BCC study resulted in approval of vismodegib, a Hedgehog pathway inhibitor indicated for treatment of adults with metastatic or locally advanced basal cell carcinoma (BCC) that has recurred after surgery or for patients who are not candidates for surgery or radiation. An efficacy and safety analysis was conducted 12 months after primary analysis. This was a multinational, multicenter, nonrandomized, 2-cohort study in patients with measurable and histologically confirmed locally advanced or metastatic BCC taking oral vismodegib (150 mg/d). Primary outcome measure was objective response rate (complete and partial responses) assessed by independent review facility. After 12 months of additional follow-up, median duration of exposure to vismodegib was 12.9 months. Objective response rate increased from 30.3% to 33.3% in patients with metastatic disease, and from 42.9% to 47.6% in patients with the locally advanced form. Median duration of response in patients with locally advanced BCC increased from 7.6 to 9.5 months. No new safety signals emerged with extended treatment duration. Limitations include low prevalence of advanced BCC and challenges of designing a study with heterogenous manifestations. The 12-month update of the study confirms the efficacy and safety of vismodegib in management of advanced BCC. Copyright © 2015 American Academy of Dermatology, Inc. Published by Elsevier Inc. All rights reserved.

Editorial: Advances in healthcare provider and patient training to improve the quality and safety of patient care

OpenAIRE

Elizabeth M. Borycki

2015-01-01

This special issue of the Knowledge Management & E-Learning: An International Journal is dedicated to describing “Advances in Healthcare Provider and Patient Training to Improve the Quality and Safety of Patient Care.” Patient safety is an important and fundamental requirement of ensuring the quality of patient care. Training and education has been identified as a key to improving healthcare provider patient safety competencies especially when working with new technologies such as electronic ...

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

International Nuclear Information System (INIS)

Tomberlin, T.A.

2002-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed

Safety design features for current UK advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

1981-01-15

The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

Safety design features for current UK advanced gas-cooled reactors

International Nuclear Information System (INIS)

Yellowlees, J.M.; Cobb, E.C.

1981-01-01

The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

Scale of production and implementation of food safety programs influence the performance of current food safety management systems: Case of dairy processors

DEFF Research Database (Denmark)

Njage, Patrick Murigu Kamau; Opiyo, Beatrice; Wangoh, John

2017-01-01

An FSMS-Diagnostic Instrument was used to evaluate fifteen Kenyan dairy processors based on indicators and descriptive grids for context riskiness, FSMS activities, and microbial food safety (FS) output with respect to scale of production. Contextual riskiness was diagnosed as low, moderate or hi....... FSMS activities were diagnosed as absent, basic, average or advanced. FS output was diagnosed as not performed, poor, moderate or good. Four clusters with significantly different (p ...

Digital instrumentation system for nuclear research reactors

International Nuclear Information System (INIS)

Aghina, Mauricio A.C.; Carvalho, Paulo Vitor R.

2002-01-01

This work describes a proposal for a system of nuclear instrumentation and safety totally digital for the Argonauta Reactor. The system divides in the subsystems: channel of pulses, channel of current, conventional instrumentation and safety system. The connection of the subsystems is made through redundant double local net, using the protocol modbus/rtu. So much the channel of pulses, the current channel and safety's system use modules operating in triple redundancy. (author)

Safety related terms for advanced nuclear plants; Terminos relacionados con la seguridad para centrales nucleares avanzadas

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety.

Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

International Nuclear Information System (INIS)

Duckwitz, Noel

2011-01-01

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets,' safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, 'Facility Safety,' and the expectations of DOE-STD-1189-2008, 'Integration of Safety into the Design Process,' provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

NRC research on the application of advanced I and C technology to commercial nuclear power plants

International Nuclear Information System (INIS)

Gollei, K.R.; Hon, A.L.

1983-01-01

The operational safety and efficiency of commercial nuclear power plants (NPP's) could possibly be enhanced by utilizing advanced instrumentation and control technology developed by other industries. The NRC is interested in learning about new I and C technology that probably will or could be applied to new or existing plants. This would enable the NRC to be better prepared to evaluate the application without undue delays. It would also help identify any appropriate changes in NRC regulations or guidance necessary to facilitate the application of advanced IandC technology to NPP's. The NRC has initiated a project to work cooperatively with the advanced technology industry, power industry, EPRI, and technical organizations such as ISA toward this goal. This paper describes the objectives and plans of this cooperative effort. It summarizes the highlights of some of the advanced technology already being evaluated by NRC such as microprocessor applications, instruments to detect inadequate core cooling and other two-phase flow measurements, reactor noise surveillance and diagnostic techniques. This paper also suggests potential candidates for consideration such as utilization of advanced instruments for LOCA experiments. It also identifies some of the potential challenges facing the application of advanced technology to NPP's. It concludes that close cooperation between NRC and industry is essential for the success of such applications

Advances in SPECT Instrumentation (Including Small Animal Scanners). Chapter 4

International Nuclear Information System (INIS)

Di Domenico, G.; Zavattini, G.

2009-01-01

Fundamental major efforts have been devoted to the development of positron emission tomography (PET) imaging modality over the last few decades. Recently, a novel surge of interest in single photon emission computed tomography (SPECT) technology has occurred, particularly after the introduction of the hybrid SPECT-CT imaging system. This has led to a flourishing of investigations in new types of detectors and collimators, and to more accurate refinement of reconstruction algorithms. Along with SPECT-CT, new, fast gamma cameras have been developed for dedicated cardiac imaging. The existing gap between PET and SPECT in sensitivity and spatial resolution is progressively decreasing, and this trend is particularly apparent in the field of small animal imaging where the most important advances have been reported in SPECT tomographs. An outline of the basic features of SPECT technology, and of recent developments in SPECT instrumentation for both clinical applications and basic biological research on animal models is described. (author)

Methodology and development of instruments for the safety analysis of a nuclear reprocessing plant

International Nuclear Information System (INIS)

Markett, J.

1987-01-01

Characteristics and overlapping aspects in the elaboration of safety analyses for the nuclear and conventional units are presented. The current methods are presented and their limits of applicability characterized. The transferability of individual methods or their elements to the analysis of the reference plant of Wackersdorf is examined and the procedure for the systems analysis is determined. It is of great importance to prove that the essential kinds of incidents and possibilities of release with potential effects in the environment are completely identified. The incidents are divided into basic incidents, which are characterized by superior physical/chemical release mechanisms. An essential objective is to systematize the safety analysis and to summarize the presentation of results. Selection criteria are presented, which allow a limitation of the analysis to essential influencing parameters without removing aspects from the overall safety-relevant statement. Besides the selection criteria, instruments and mathematical models are explained with the help of which the representative and possible incidents covering all potential risks for all areas of the plant, systems and components can be selected. These design-basis accidents (criticality, self-heating, fire, explosion, leakages, earth quakes) are decisive for the determination of potential damaging effects in the environment and thus for the overall statement on the licensability. (orig./HP) [de

Instrumentation

International Nuclear Information System (INIS)

Decreton, M.

2002-01-01

SCK-CEN's R and D programme on instrumentation involves the development of advanced instrumentation systems for nuclear applications as well as the assessment of the performance of these instruments in a radiation environment. Particular emphasis is on the use of optical fibres as umbilincal links of a remote handling unit for use during maintanance of a fusion reacor, studies on the radiation hardening of plasma diagnostic systems; investigations on new instrumentation for the future MYRRHA accelerator driven system; space applications related to radiation-hardened lenses; the development of new approaches for dose, temperature and strain measurements; the assessment of radiation-hardened sensors and motors for remote handling tasks and studies of dose measurement systems including the use of optical fibres. Progress and achievements in these areas for 2001 are described

Instrumentation

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2002-04-01

SCK-CEN's R and D programme on instrumentation involves the development of advanced instrumentation systems for nuclear applications as well as the assessment of the performance of these instruments in a radiation environment. Particular emphasis is on the use of optical fibres as umbilincal links of a remote handling unit for use during maintanance of a fusion reacor, studies on the radiation hardening of plasma diagnostic systems; investigations on new instrumentation for the future MYRRHA accelerator driven system; space applications related to radiation-hardened lenses; the development of new approaches for dose, temperature and strain measurements; the assessment of radiation-hardened sensors and motors for remote handling tasks and studies of dose measurement systems including the use of optical fibres. Progress and achievements in these areas for 2001 are described.

Heat Flux Instrumentation Laboratory (HFIL)

Data.gov (United States)

Federal Laboratory Consortium — Description: The Heat Flux Instrumentation Laboratory is used to develop advanced, flexible, thin film gauge instrumentation for the Air Force Research Laboratory....

The Gas Turbine Modular Helium Reactor Optimum Design for Instrumentation and Control

International Nuclear Information System (INIS)

Quinn, Edward 'Ted'; Rodriguez, Carmelo

2002-01-01

Nuclear power plants rely on instrumentation and control (I and C) systems for monitoring, control and protection. While the earlier generations of nuclear plants were built with largely analog instrumentation systems, the industrial base has moved to digital based system. The reason for this transition to digital I and C systems lies in their important advantage over existing analog systems. In essence, the challenge is to prove that the new digital guidelines included in NUREG 0800, Standard Review Plan Chapter 7 and all associated documents provided adequate direction and safety review for an advanced control design for a new nuclear plant in the United States. This paper focuses on how one of the new designs incorporates the latest technology and meets the new NRC requirements. (authors)

Safety climate and attitude as evaluation measures of organizational safety.

Science.gov (United States)

Isla Díaz, R; Díaz Cabrera, D

1997-09-01

The main aim of this research is to develop a set of evaluation measures for safety attitudes and safety climate. Specifically it is intended: (a) to test the instruments; (b) to identify the essential dimensions of the safety climate in the airport ground handling companies; (c) to assess the quality of the differences in the safety climate for each company and its relation to the accident rate; (d) to analyse the relationship between attitudes and safety climate; and (e) to evaluate the influences of situational and personal factors on both safety climate and attitude. The study sample consisted of 166 subjects from three airport companies. Specifically, this research was centered on ground handling departments. The factor analysis of the safety climate instrument resulted in six factors which explained 69.8% of the total variance. We found significant differences in safety attitudes and climate in relation to type of enterprise.

Contribution to the safety assessment of instrumentation and control software for nuclear power plants. Application to spin N4

International Nuclear Information System (INIS)

Soubies, B.; Boulc'h, J.; Elsensohn, O.; Le Meur, M.; Henry, J.Y.

1994-01-01

The process of licensing nuclear power plants for operation consists of mandatory steps featuring detailed examination of the instrumentation and control system. Significant changes were introduced by the operator in the process of designing and producing 1400 MWe pressurized water reactor safety systems and, in particular, in the case of the Digital Integrated Protection System, (French abbreviation SPIN). The methodology applied by the Institute of Protection and Nuclear Safety (IPSN) to examine the software of this system is described. It consists of the methods used by the manufacturer to develop SPIN software for the 1400 MWe PWRs, and the approach adopted by the IPSN to evaluate SPIN safety softwares of the protection system for the N4 series of reactors. (R.P.). 2 refs

The development of advanced instrumentation and control technology. The development of verification and validation technology for instrumentation and control in NPPs

International Nuclear Information System (INIS)

Kwon, Kee Choon; Ham, Chang Shik; Lee, Jang Soo; Kim, Jang Yeol; Song, Soon Ja; Kim, Jung Taek; Park, Won Man; Lee, Dong Young; Eom, Heung Seop; Park, Kee Yong.

1997-07-01

We found essential problems followed by digitalizing of instrumentation and control. A scheme is divided into hardware and software to resolve these problems. We have analyzed the hardware V and V methodologies about common mode failure, commercial grade dedication process and electromagnetic compatibility. We have developed several guidelines, the software classification guideline, the quality assurance handbook for the software in digital I and C, the software V and V planning guideline, and the software safety guideline. And then, we have established the integrated environment for the safety-critical software based on the Computer Aided Software Engineering (CASE) tools. We have also surveyed a trend and application case of test facility for establishment of functional requirements. Input/output interface which connects among the host computer and developed target and panel are designed and manufactured using UXI bus. The developed functional test facility is used for test and validate the automatic start-up intelligent control system, the dynamic alarm system, the accident identification system using hidden Markov model, and the intelligent logic tracking system. The result of evaluation of the above systems shows the functional test facility performance is sufficient in normal operating and transient conditions. (author). 24 tabs., 59 figs

The development of advanced instrumentation and control technology. The development of verification and validation technology for instrumentation and control in NPPs

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee Choon; Ham, Chang Shik; Lee, Jang Soo; Kim, Jang Yeol; Song, Soon Ja; Kim, Jung Taek; Park, Won Man; Lee, Dong Young; Eom, Heung Seop; Park, Kee Yong

1997-07-01

We found essential problems followed by digitalizing of instrumentation and control. A scheme is divided into hardware and software to resolve these problems. We have analyzed the hardware V and V methodologies about common mode failure, commercial grade dedication process and electromagnetic compatibility. We have developed several guidelines, the software classification guideline, the quality assurance handbook for the software in digital I and C, the software V and V planning guideline, and the software safety guideline. And then, we have established the integrated environment for the safety-critical software based on the Computer Aided Software Engineering (CASE) tools. We have also surveyed a trend and application case of test facility for establishment of functional requirements. Input/output interface which connects among the host computer and developed target and panel are designed and manufactured using UXI bus. The developed functional test facility is used for test and validate the automatic start-up intelligent control system, the dynamic alarm system, the accident identification system using hidden Markov model, and the intelligent logic tracking system. The result of evaluation of the above systems shows the functional test facility performance is sufficient in normal operating and transient conditions. (author). 24 tabs., 59 figs.

Nanopositioning techniques development for synchrotron radiation instrumentation applications at the Advanced Photon Source

International Nuclear Information System (INIS)

Shu Deming

2010-01-01

At modern synchrotron radiation sources and beamlines, high-precision positioning techniques present a significant opportunity to support state-of-the-art synchrotron radiation research. Meanwhile, the required instrument positioning performance and capabilities, such as resolution, dynamic range, repeatability, speed, and multiple axes synchronization are exceeding the limit of commercial availability. This paper presents the current nanopositioning techniques developed for the Argonne Center for Nanoscale Materials (CNM)/Advanced Photon Source (APS) hard x-ray nanoprobe and high-resolution x-ray monochromators and analyzers for the APS X-ray Operations and Research (XOR) beamlines. Future nanopositioning techniques to be developed for the APS renewal project will also be discussed.

Euro NCAP, a safety instrument.

NARCIS (Netherlands)

2007-01-01

Since early 1997, in Europe, a cooperation of consumer organizations, European governments, the European Commission, and car organizations have been doing crash tests to judge the safety of cars. This programme is called Euro NCAP. The test programme consists of four tests of new cars to see how

SWR 1000: an advanced boiling water reactor with passive safety features

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

The Impact of Advanced Age on Driving Safety in Adults with Medical Conditions.

Science.gov (United States)

Moon, Sanghee; Ranchet, Maud; Akinwuntan, Abiodun Emmanuel; Tant, Mark; Carr, David Brian; Raji, Mukaila Ajiboye; Devos, Hannes

2018-01-01

Adults aged 85 and older, often referred to as the oldest-old, are the fastest-growing segment of the population. The rapidly increasing number of older adults with chronic and multiple medical conditions poses challenges regarding their driving safety. To investigate the effect of advanced age on driving safety in drivers with medical conditions. We categorized 3,425 drivers with preexisting medical conditions into four age groups: middle-aged (55-64 years, n = 1,386), young-old (65-74 years, n = 1,013), old-old (75-84 years, n = 803), or oldest-old (85 years and older, n = 223). All underwent a formal driving evaluation. The outcome measures included fitness to drive recommendation by the referring physician, comprehensive fitness to drive decision from an official driving evaluation center, history of motor vehicle crashes (MVCs), and history of traffic violations. The oldest-old reported more cardiopulmonary and visual conditions, but less neurological conditions than the old-old. Compared to the middle-aged, the oldest-old were more likely to be considered unfit to drive by the referring physicians (odds ratio [OR] = 4.47, 95% confidence interval [CI] 2.20-9.10) and by the official driving evaluation center (OR = 2.74, 95% CI 1.87-4.03). The oldest-old reported more MVCs (OR = 2.79, 95% CI 1.88-4.12) compared to the middle-aged. Advanced age adversely affected driving safety outcomes. The oldest-old are a unique age group with medical conditions known to interfere with safe driving. Driving safety strategies should particularly target the oldest-old since they are the fastest-growing group and their increased frailty is associated with severe or fatal injuries due to MVCs. © 2018 S. Karger AG, Basel.

The Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management. An instrument to achieve a global safety

International Nuclear Information System (INIS)

Risoluti, P.

2006-01-01

The Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management (the Joint Convention) is the first legally binding international treaty in the area of radioactive material management. It was adopted by a Diplomatic Conference in September 1997 and opened for signature on 29 September 1997. The Convention entered into force on 18 June 1998, and to date (May 2006) has been ratified by 41 countries. The Joint Convention applies to spent fuel and radioactive waste resulting from civilian application. Its principal aim is to achieve and maintain a high degree of safety in their management worldwide. The Convention is an incentive instrument, not designed to ensure fulfilment of obligations through control and sanction, but by a volunteer peer review mechanism. The obligations of the Contracting Parties are mainly based on the international safety standards developed by the IAEA in past decades. The Convention is of interest of all countries generating radioactive waste. Therefore it is relevant not only for those using nuclear power, but for any country where application of nuclear energy in education, agriculture, medicine and industry is currently used. Obligations of Contracting Parties include attending a Review Meeting held every three years and prepare National Reports for review by the other Contracting Parties. In the National Reports basic information on inventory and facilities for management of radioactive materials has to be provided. Countries with small nuclear power and/or research programs or countries having radioactive materials only from nuclear application on medicine, agriculture or conventional industry, can benefit from the exchange of information and the technical knowledge gained by the reporting procedure set up by the Convention. The second Review Meeting is to be held at IAEA headquarters from 15 to 26 May 2006. This paper presents the objectives and the implementation status of the Convention, the

System safety education focused on flight safety

Science.gov (United States)

Holt, E.

1971-01-01

The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.

AREVA advanced safety IC solutions and licensing experience for new nuclear builds and modernization projects - 15545

International Nuclear Information System (INIS)

Fourestie, B.; Pickelmann, J.; Richter, S.; Hilsenkopf, P.; Paris, P.

2015-01-01

Regulatory requirements for the Instrumentation and Control (IC) for Nuclear Power Plants have become significantly more stringent during the last 10 years in the areas of software development and qualification, traceability, diversity, or seismic requirements for instance, and with the introduction of new standards (such as the IEC 62566, or the IEC 62003). Based on a large and comprehensive experience gained from projects in several regulatory environments and different plant types (including non-OEM plants), AREVA has developed and adapted its processes and products to provide state-of-the-art IC solutions in full compliance with the regulatory demands and requirements in terms of robustness (independence, defense-in-depth, diversity and cyber-security). In this paper we present the safety IC platforms developed by AREVA. These platforms include TELEPERM XS as the computerized safety IC platform for class 1 system implementation, the Qualified Display System (QDS) for safety classified screen-based interface, and UNICORN as fully diverse analog safety IC platform for backup systems

Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.1

International Nuclear Information System (INIS)

2014-01-01

The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.2

International Nuclear Information System (INIS)

2014-01-01

The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

International Nuclear Information System (INIS)

Wootan, D.W.; Omberg, R.P.

2015-01-01

Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

FY09 Advanced Instrumentation and Active Interrogation Research for Safeguards

International Nuclear Information System (INIS)

Chichester, D.L.; Pozzi, S.A.; Seabury, E.H.; Dolan, J.L.; Flaska, M.; Johnson, J.T.; Watson, S.M.; Wharton, J.

2009-01-01

Multiple small-scale projects have been undertaken to investigate advanced instrumentation solutions for safeguard measurement challenges associated with advanced fuel cycle facilities and next-generation fuel reprocessing installations. These activities are in support of the U.S. Department of Energy's Fuel Cycle Research and Development program and its Materials Protection, Accounting, and Control for Transmutation (MPACT) campaign. (1) Work was performed in a collaboration with the University of Michigan (Prof. Sara Pozzi, co-PI) to investigate the use of liquid-scintillator radiation detectors for assaying mixed-oxide (MOX) fuel, to characterize its composition and to develop advanced digital pulse-shape discrimination algorithms for performing time-correlation measurements in the MOX fuel environment. This work included both simulations and experiments and has shown that these techniques may provide a valuable approach for use within advanced safeguard measurement scenarios. (2) Work was conducted in a collaboration with Oak Ridge National Laboratory (Dr. Paul Hausladen, co-PI) to evaluate the strengths and weaknesses of the fast-neutron coded-aperture imaging technique for locating and characterizing fissile material, and as a tool for performing hold-up measurements in fissile material handling facilities. This work involved experiments at Idaho National Laboratory, using MOX fuel and uranium metal, in both passive and active interrogation configurations. A complete analysis has not yet been completed but preliminary results suggest several potential uses for the fast neutron imaging technique. (3) Work was carried out to identify measurement approaches for determining nitric acid concentration in the range of 1-4 M and beyond. This work included laboratory measurements to investigate the suitability of prompt-gamma neutron activation analysis for this measurement and product reviews of other commercial solutions. Ultrasonic density analysis appears to be

Obtaining laser safety at a synchrotron radiation user facility: The Advanced Light Source

International Nuclear Information System (INIS)

Barat, K.

1996-01-01

The Advanced Light Source (ALS) is a US national facility for scientific research and development located at the Lawrence Berkeley National Laboratory in California. The ALS delivers the world's brightest synchrotron radiation in the far ultraviolet and soft X-ray regions of the spectrum. As a user facility it is available to researchers from industry, academia, and laboratories from around the world. Subsequently, a wide range of safety concerns become involved. This article relates not only to synchrotron facilities but to any user facility. A growing number of US centers are attracting organizations and individuals to use the equipment on site, for a fee. This includes synchrotron radiation and/or free electron facilities, specialty research centers, and laser job shops. Personnel coming to such a facility bring with them a broad spectrum of safety cultures. Upon entering, the guests must accommodate to the host facility safety procedures. This article describes a successful method to deal with that responsibility

Neutron-multiplication measurement instrument

Energy Technology Data Exchange (ETDEWEB)

Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

1982-01-01

The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results.

Neutron multiplication measurement instrument

International Nuclear Information System (INIS)

Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

1983-01-01

The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results

Neutron-multiplication measurement instrument

International Nuclear Information System (INIS)

Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

1982-01-01

The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results

Present scenery of cuban legislation in the field of legal verification of dosimetric instruments used in radiological protection

International Nuclear Information System (INIS)

Salas G, Walwyn; Morales Monzon, J.A.; Hernandez Blanche, E.

2001-01-01

The main objective of legal metrology is to ensure the public guaranty from the point of view of safety, and the suitable accuracy of the measurements that are made on health, environmental applications, and trade. The International Organization of Legal Metrology included the ionizing radiation field on those for which the use of the verified measuring instruments are suggested. . The paper presents the advances of Cuban legislation in this field, promoted by issue of the Decree-Law 183 of Metrology. As part of such advances, the Cuban standards for verification NC 44:1999 'X and Gamma Radiation Measuring Instruments. Verification methods' is discussed. This standard was elaborated in the Cuban Secondary Standard Dosimetry Laboratory, and it is based on the available relevant international standards. Results from verification service during the year 2000 are also provided.(author)

Application of Advanced Technology to Improve Plant Performance. Safety and Performance in Current NPPs

International Nuclear Information System (INIS)

Hashemian, H.M.

2011-01-01

Advances in computer technologies, signal processing, analytical modeling, and the advent of wireless sensors have provided the nuclear industry with ample means to automate and optimize maintenance activities and improve safety, efficiency, and availability, while reducing costs and radiation exposure to maintenance personnel. This paper provides a review of these developments and presents examples of their use in the nuclear power industry and the financial and safety benefits that they have produced. As the current generation of nuclear power plants have passed their mid-life, increased monitoring of their health is critical to their safe operation. This is especially true now that license renewal of nuclear power plants has accelerated, allowing some plants to operate up to 60 years or more. Furthermore, many utilities are maximizing their power output through uprating projects and retrofits. This puts additional demand and more stress on the plant equipment such as the instrumentation and control (I and C) systems and the reactor internal components making them more vulnerable to the effects of aging, degradation, and failure. In the meantime, the nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automation of testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. The foundation for much of the required technologies has already been established through 40 years of research and development (R and D) efforts performed by numerous organizations, scientists, and engineers around the world including the author. This paper provides examples of these technologies and demonstrates how the gap between some of the more important R and D efforts and end users have been filled

Design concepts for a nuclear digital instrumentation and control system platform

International Nuclear Information System (INIS)

Ou, T. C.; Chen, C. K.; Chen, P. J.; Shyu, S. S.; Lee, C. L.; Hsieh, S. F.

2010-10-01

The objective of this paper is to present the development results of the nuclear instrumentation and control system in Taiwan. As the Taiwan nuclear power plants age, the need to consider upgrading of both their safety and non-safety-related instrumentation and control systems becomes more urgent. Meanwhile, the digital instrumentation and control system that is based on current fast evolving electronic and information technologies are difficult to maintain effectively. Therefore, Institute of Nuclear Energy Research was made a decision to promote the Taiwan Nuclear Instrumentation and Control System project to collaborate with domestic electronic industry to establish self-reliant capabilities on the design, manufacturing, and application of nuclear instrumentation and control systems with newer technology. In the case of safety-related applications like nuclear instrumentation and control, safety-oriented quality control is required. In order to establish a generic qualified digital platform, the world-wide licensing experience should be considered in the licensing process. This paper describes the qualification and certification tools by IEC 61508 for design and development of safety related equipment and explains the basis for many decisions made while performing the digital upgrade. (Author)

Comparative instrumental evaluation of efficacy and safety between a binary and a ternary system in chemexfoliation.

Science.gov (United States)

Cameli, Norma; Mariano, Maria; Ardigò, Marco; Corato, Cristina; De Paoli, Gianfranco; Berardesca, Enzo

2017-09-20

To instrumentally evaluate the efficacy and the safety of a new ternary system chemo exfoliating formulation (water-dimethyl isosorbide-acid) vs traditional binary systems (water and acid) where the acid is maintained in both the systems at the same concentration. Different peelings (binary system pyruvic acid and trichloroacetic acid-TCA, and ternary system pyruvic acid and TCA) were tested on the volar forearm of 20 volunteers of both sexes between 28 and 50 years old. The outcomes were evaluated at the baseline, 10 minutes, 24 hours, and 1 week after the peeling by means of noninvasive skin diagnosis techniques. In vivo reflectance confocal microscopy was used for stratum corneum evaluation, transepidermal waterloss, and Corneometry for skin barrier and hydration, Laser Doppler velocimetry in association with colorimetry for irritation and erythema analysis. The instrumental data obtained showed that the efficacy and safety of the new ternary system peel compounds were significantly higher compared with the binary system formulations tested. The new formulation peels improved chemexfoliation and reduced complications such as irritation, redness, and postinflammatory pigmentation compared to the traditional aqueous solutions. The study showed that ternary system chemexfoliation, using a controlled delivery technology, was able to provide the same clinical effects in term of stratum corneum reduction with a significantly reduced barrier alteration, water loss, and irritation/erythema compared to traditional binary system peels. © 2017 Wiley Periodicals, Inc.

Reliability prediction for the vehicles equipped with advanced driver assistance systems (ADAS and passive safety systems (PSS

Directory of Open Access Journals (Sweden)

Balbir S. Dhillon

2012-10-01

Full Text Available The human error has been reported as a major root cause in road accidents in today’s world. The human as a driver in road vehicles composed of human, mechanical and electrical components is constantly exposed to changing surroundings (e.g., road conditions, environmentwhich deteriorate the driver’s capacities leading to a potential accident. The auto industries and transportation authorities have realized that similar to other complex and safety sensitive transportation systems, the road vehicles need to rely on both advanced technologies (i.e., Advanced Driver Assistance Systems (ADAS and Passive Safety Systems (PSS (e.g.,, seatbelts, airbags in order to mitigate the risk of accidents and casualties. In this study, the advantages and disadvantages of ADAS as active safety systems as well as passive safety systems in road vehicles have been discussed. Also, this study proposes models that analyze the interactions between human as a driver and ADAS Warning and Crash Avoidance Systems and PSS in the design of vehicles. Thereafter, the mathematical models have been developed to make reliability prediction at any given time on the road transportation for vehicles equipped with ADAS and PSS. Finally, the implications of this study in the improvement of vehicle designs and prevention of casualties are discussed.

Technology Roadmap Instrumentation, Control, and Human-Machine Interface to Support DOE Advanced Nuclear Energy Programs

Energy Technology Data Exchange (ETDEWEB)

Donald D Dudenhoeffer; Burce P Hallbert

2007-03-01

Instrumentation, Controls, and Human-Machine Interface (ICHMI) technologies are essential to ensuring delivery and effective operation of optimized advanced Generation IV (Gen IV) nuclear energy systems. In 1996, the Watts Bar I nuclear power plant in Tennessee was the last U.S. nuclear power plant to go on line. It was, in fact, built based on pre-1990 technology. Since this last U.S. nuclear power plant was designed, there have been major advances in the field of ICHMI systems. Computer technology employed in other industries has advanced dramatically, and computing systems are now replaced every few years as they become functionally obsolete. Functional obsolescence occurs when newer, more functional technology replaces or supersedes an existing technology, even though an existing technology may well be in working order.Although ICHMI architectures are comprised of much of the same technology, they have not been updated nearly as often in the nuclear power industry. For example, some newer Personal Digital Assistants (PDAs) or handheld computers may, in fact, have more functionality than the 1996 computer control system at the Watts Bar I plant. This illustrates the need to transition and upgrade current nuclear power plant ICHMI technologies.

Advances in luminescence instrument systems

DEFF Research Database (Denmark)

Bøtter-Jensen, L.; Bulur, E.; Duller, G.A.T.

2000-01-01

We report on recent advances in the development of luminescence measurement systems and techniques at Riso. These include: (1) optical stimulation units based on new-generation powerful blue light (470 nm) emitting diodes providing up to 28 mW/cm(2) for OSL measurements; (2) an infrared (830 nm...

24 CFR 232.570 - Endorsement of credit instrument.

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 2 2010-04-01 2010-04-01 false Endorsement of credit instrument... of Fire Safety Equipment Eligible Security Instruments § 232.570 Endorsement of credit instrument. The Commissioner shall indicate his insurance of the loan by endorsing the credit instrument and...

Technical Support Section Instrument Support Program for nuclear and nonnuclear facilities with safety requirements

International Nuclear Information System (INIS)

Adkisson, B.P.; Allison, K.L.

1995-01-01

This document describes requirements, procedures, and supervisory responsibilities of the Oak Ridge National Laboratory (ORNL) Instrumentation and Controls (I ampersand C) Division's Technical Support Section (TSS) for instrument surveillance and maintenance in nonreactor nuclear facilities having identified Operational Safety Requirements (OSRs) or Limiting Conditions Document (LCDs). Implementation of requirements comply with the requirements of U.S. Department of Energy (DOE) Orders 5480.5, 5480.22, and 5481.1B; Martin Marietta Energy Systems, Inc. (Energy Systems), Policy Procedure ESS-FS-201; and ORNL SPP X-ESH-15. OSRs and LCDs constitute an agreement or contract between DOE and the facility operating management regarding the safe operation of the facility. One basic difference between OSRs and LCDs is that violation of an OSR is considered a Category II occurrence, whereas violation of an LCD requirement is considered a Category III occurrence (see Energy Systems Standard ESS-OP-301 and ORNL SPP X-GP-13). OSRs are required for high- and moderate-hazard nuclear facilities, whereas the less-rigorous LCDs are required for low-hazard nuclear facilities and selected open-quotes generally acceptedclose quotes operations. Hazard classifications are determined through a hazard screening process, which each division conducts for its facilities

Proceedings of the 5. International Topical Meeting on Nuclear Plant Instrumentation Controls, and Human Machine Interface Technology

International Nuclear Information System (INIS)

2006-01-01

Instrumentation, Controls, and Human-Machine Interface (ICHMI) technologies are essential to ensuring delivery and effective operation of nuclear power systems. The ICHMI system, together with plant personnel, is the 'central nervous system' for operating plants. It senses basic parameters, monitors performance, integrates information, and makes adjustments to plant operations as necessary. It also responds to failures and off-normal events, thus ensuring goals of efficient power production and safety. The ICHMI system embodies the sensing, communications, monitoring, control, and presentation and command systems between the process (i.e., the reactor, heat transport, and energy conversion systems) and the plant personnel. It enables plant personnel to more effectively monitor the health of the plant and to identify opportunities to improve the performance of equipment and systems as well as to anticipate, understand, and respond to potential problems. Improved controls provide the basis to operate more closely to performance margins, and the improved integration of automatic and human response enables them to work cooperatively to accomplish production and safety goals. The ICHMI system thus directly impacts the performance of the entire plant and thereby the economics, safety, and security of current and future reactor designs. The 5. International Topical Meeting on Nuclear Plant Instrumentation Control and Human-Machine Interface Technology (NPIC and HMIT 2006) is specifically devoted to advances in these important technologies. In these proceedings, more than 200 papers and panel sessions from all over the world have been assembled to share the most recent information and innovations in ICHMI technology and to discuss the important issues that face the future of the industry. The papers fall into two major groupings: instrumentation and control (I and C) and human-machine interface technology (HMIT). The I and C papers are organized into five tracks. 'Systems

Proceedings of the 5. International Topical Meeting on Nuclear Plant Instrumentation Controls, and Human Machine Interface Technology

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

Instrumentation, Controls, and Human-Machine Interface (ICHMI) technologies are essential to ensuring delivery and effective operation of nuclear power systems. The ICHMI system, together with plant personnel, is the 'central nervous system' for operating plants. It senses basic parameters, monitors performance, integrates information, and makes adjustments to plant operations as necessary. It also responds to failures and off-normal events, thus ensuring goals of efficient power production and safety. The ICHMI system embodies the sensing, communications, monitoring, control, and presentation and command systems between the process (i.e., the reactor, heat transport, and energy conversion systems) and the plant personnel. It enables plant personnel to more effectively monitor the health of the plant and to identify opportunities to improve the performance of equipment and systems as well as to anticipate, understand, and respond to potential problems. Improved controls provide the basis to operate more closely to performance margins, and the improved integration of automatic and human response enables them to work cooperatively to accomplish production and safety goals. The ICHMI system thus directly impacts the performance of the entire plant and thereby the economics, safety, and security of current and future reactor designs. The 5. International Topical Meeting on Nuclear Plant Instrumentation Control and Human-Machine Interface Technology (NPIC and HMIT 2006) is specifically devoted to advances in these important technologies. In these proceedings, more than 200 papers and panel sessions from all over the world have been assembled to share the most recent information and innovations in ICHMI technology and to discuss the important issues that face the future of the industry. The papers fall into two major groupings: instrumentation and control (I and C) and human-machine interface technology (HMIT). The I and C papers are organized into five tracks

Advanced human-system interface design review guidelines

International Nuclear Information System (INIS)

O'Hara, J.M.

1990-01-01

Advanced, computer-based, human-system interface designs are emerging in nuclear power plant control rooms as a result of several factors. These include: (1) incorporation of new systems such as safety parameter display systems, (2) backfitting of current control rooms with new technologies when existing hardware is no longer supported by equipment vendors, and (3) development of advanced control room concepts. Control rooms of the future will be developed almost exclusively with advanced instrumentation and controls based upon digital technology. In addition, the control room operator will be interfacing with more intelligent systems which will be capable of providing information processing support to the operator. These developments may have significant implications for plant safety in that they will greatly affect the operator's role in the system as well as the ways in which he interacts with it. At present, however, the only guidance available to the Nuclear Regulatory Commission (NRC) for the review of control room-operator interfaces is NUREG-0700. It is a document which was written prior to these technological changes and is, therefore, tailored to the technologies used in traditional control rooms. Thus, the present guidance needs to be updated since it is inadequate to serve as the basis for NRC staff review of such advanced or hybrid control room designs. The objective of the project reported in this paper is to develop an Advanced Control Room Design Review Guideline suitable for use in performing human factors reviews of advanced operator interfaces. This guideline will take the form of a portable, interactive, computer-based document that may be conveniently used by an inspector in the field, as well as a text-based document

Advances in nuclear medicine instrumentation: considerations in the design and selection of an imaging system

International Nuclear Information System (INIS)

Links, J.M.

1998-01-01

Nuclear medicine remains a vibrant and dynamic medical specialty because it so adeptly marries advances in basic science research, technology, and medical practice in attempting to solve patients' problems. As a physicist, it is my responsibility to identify or design new instrumentation and techniques, and to implement, validate, and help apply these new approaches in the practice of nuclear medicine. At Johns Hopkins, we are currently in the process of purchasing both a single-photon/coincidence tomographic imaging system and a dedicated positron emission tomography (PET) scanner. Given the exciting advances that have been made, but the conflicting opinions of manufacturers and colleagues alike regarding ''best'' choices, it seemed useful to review what is new now, and what is on the horizon, to help identify all of the important considerations in the design and selection of an imaging system. It is important to note that many of the ''advances'' described here are in an early stage of development, and may never make it to routine clinical practice. Further, not all of the advances are of equal importance, or have the same degree of general clinical applicability. Please also note that the references contained herein are for illustrative purposes and are not all-inclusive; no implication that those chosen are ''better'' than others not mentioned is intended. (orig.)

Development of Interactive Monitoring System for Neutron Scattering Instrument

Energy Technology Data Exchange (ETDEWEB)

So, Ji Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Neutron scattering instruments in HANARO research reactor have been contributed to various fields of basic science and material engineering. These instruments are open to publics and researchers can apply beam-time and do experiments with instrument scientists. In most cases, these instruments run for several weeks without stopping, and therefore instrument scientist wants to see the instrument status and receive information if the instruments have some problem. This is important for the safety. However, it is very hard to get instrument information outside of instruments. Access from external site is strongly forbidden in the institute due to the network safety, I developed another way to send instrument status information using commercial short messaging service(SMS). In this presentation, detailed features of this system will be shown. As a prototype, this system is being developed for the single instrument: Disk-chopper time-of-flight instruments (DC-TOF). I have successfully developed instruments and operate for several years. This information messaging system can be used for other neutron scattering instruments.

Performance of food safety management systems in poultry meat preparation processing plants in relation to Campylobacter spp. contamination.

Science.gov (United States)

Sampers, Imca; Jacxsens, Liesbeth; Luning, Pieternel A; Marcelis, Willem J; Dumoulin, Ann; Uyttendaele, Mieke

2010-08-01

A diagnostic instrument comprising a combined assessment of core control and assurance activities and a microbial assessment instrument were used to measure the performance of current food safety management systems (FSMSs) of two poultry meat preparation companies. The high risk status of the company's contextual factors, i.e., starting from raw materials (poultry carcasses) with possible high numbers and prevalence of pathogens such as Campylobacter spp., requires advanced core control and assurance activities in the FSMS to guarantee food safety. The level of the core FSMS activities differed between the companies, and this difference was reflected in overall microbial quality (mesophilic aerobic count), presence of hygiene indicators (Enterobacteriaceae, Staphylococcus aureus, and Escherichia coli), and contamination with pathogens such as Salmonella, Listeria monocytogenes, and Campylobacter spp. The food safety output expressed as a microbial safety profile was related to the variability in the prevalence and contamination levels of Campylobacter spp. in poultry meat preparations found in a Belgian nationwide study. Although a poultry meat processing company could have an advanced FSMS in place and a good microbial profile (i.e., lower prevalence of pathogens, lower microbial numbers, and less variability in microbial contamination), these positive factors might not guarantee pathogen-free products. Contamination could be attributed to the inability to apply effective interventions to reduce or eliminate pathogens in the production chain of (raw) poultry meat preparations.

Thermal Testing and Model Correlation for Advanced Topographic Laser Altimeter Instrument (ATLAS)

Science.gov (United States)

Patel, Deepak

2016-01-01

The Advanced Topographic Laser Altimeter System (ATLAS) part of the Ice Cloud and Land Elevation Satellite 2 (ICESat-2) is an upcoming Earth Science mission focusing on the effects of climate change. The flight instrument passed all environmental testing at GSFC (Goddard Space Flight Center) and is now ready to be shipped to the spacecraft vendor for integration and testing. This topic covers the analysis leading up to the test setup for ATLAS thermal testing as well as model correlation to flight predictions. Test setup analysis section will include areas where ATLAS could not meet flight like conditions and what were the limitations. Model correlation section will walk through changes that had to be made to the thermal model in order to match test results. The correlated model will then be integrated with spacecraft model for on-orbit predictions.

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

International Nuclear Information System (INIS)

2000-12-01

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

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

International Nuclear Information System (INIS)

2000-12-01

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

Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

International Nuclear Information System (INIS)

King, T.L.; Landry, R.R.; Throm, E.D.; Wilson, J.N.

1991-12-01

This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

Operational safety

International Nuclear Information System (INIS)

Anon.

1977-01-01

The PNL Safety, Standards and Compliance Program contributed to the development and issuance of safety policies, standards, and criteria; for projects in the nuclear and nonnuclear areas. During 1976 the major emphasis was on developing criteria, instruments and methods to assure that radiation exposure to occupational personnel and to people in the environs of nuclear-related facilities is maintained at the lowest level technically and economically practicable. Progress in 1976 is reported on the preparation of guidelines for radiation exposure; Pu dosimetry studies; the preparation of an environmental monitoring handbook; and emergency instrumentation preparedness

Safety

International Nuclear Information System (INIS)

2001-01-01

This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

The impact of twenty years of noise research on nuclear power plant design, instrumentation and control

International Nuclear Information System (INIS)

Kemeny, L.G.

1975-01-01

Early investigations demonstrated that time constants and the dynamic characteristics of low energy nuclear systems could be elegantly determined by correlation of spectral analysis of fluctuating signals from ion chambers and proportional counters. Analyses of the time series information and the multi-filtering operations in the frequency domain were time consuming and tedious projects due to the lack of suitable data processing equipment. During the last decade, the significant advances were the recognition of the advantages of the two-channel cross-correlation technique and the realisation that the dynamic behaviour of nuclear power plant at power could be monitored and studied in depth by the cross-correlation of mechanical, thermal and hydrodynamic signals with neutronic information. The former concept led to the development of theoretical models for spatial and energy-dependent noise fields within a nuclear system. The latter opened a floodgate of potential advances in nuclear power plant design optimization, control and safety instrumentation, and control and safety diagnostic systems. (U.K.)

Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

International Nuclear Information System (INIS)

Boeck, H.; Weiss, H.; Hood, W.E.; Hyde, W.K.

1992-01-01

The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert is assigned the responsibility of reviewing the proposed installation and verifying all safety aspects. After a positive review, final assembly and checkout of the instrumentation system may commence. The instrumentation system consists of three basic modules: the control system console, the data acquisition console and the NH-1000 wide range channel. Digital communications greatly reduce interwiring requirements. Hardwired safety channels are independent of computer control, thus, the instrumentation system in no way relies on any computer intervention for safety function. In addition, both the CSC and DAC computers are continuously monitored for proper operation via watchdog circuits which are capable of shutting down the reactor in the event of computer malfunction. Safety channels include two interlocked NMP-1000 multi-range linear channels for steady state mode, an NPP-1000 linear safety channel for pulse mode and a set of three independent fuel temperature monitoring channels. The microprocessor controlled wide range NM- 1000 digital neutron monitor (fission chamber based) functions as a startup/operational channel, and provides all power level related Interlocks. The Atominstitut TRIGA reactor is configured for four modes of operation: manual mode, automatic mode (servo control), pulsing mode and square wave mode. Control of the standard control rods is via stepping motor control rod drives, which offers the operator the choice of which control rods are operated by the servo system in automatic and square wave model. (author)

Laser Induced Breakdown Spectroscopy, advances in resolution and portability

International Nuclear Information System (INIS)

Ponce, L.; Flores, T.; Arronte, M.; Moreira, L.; Hernandez, L. C.; Posada, E. de

2009-01-01

Laser Induced Breakdown Spectroscopy (LIBS), can be considered as one of the most dynamic and promising technique in the field of analytical spectroscopy. LIBS has turned into a powerful alternative for a wide front of applications, from the geological exploration to the industrial inspection, including the environmental monitoring, the biomedical analysis, the study of patrimonial works, the safety and defense. The advances in LIBS instrumentation have allowed improving gradually the analysis services and quality, on the basis of a better knowledge of the technology principles. Recently, systems of double pulse have facilitated a better dosing of energy, the improvement of the signal-noise relation and the study of the different process stages. Femtosecond lasers offers the possibility of study in detail the ablation and atomic emission processes. New advances like multi-pulse or multi-wavelength systems -in fact stilling without exploring, must offer new information to advance in this knowledge. Finally, which it does to this technology really attractive, is the aptitude to be employed in field conditions, or for the detection of the elementary composition at long distances. In this presentation there are discussed the designs of portable instrumentation, compact and low cost, which can improve substantially the LIBS possibilities. (Author)

Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

International Nuclear Information System (INIS)

Duckwitz, Noel

2011-01-01

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets,' safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, 'Facility Safety,' and the expectations of DOE-STD-1189-2008, 'Integration of Safety into the Design Process,' provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Advanced technology for nuclear powerplants

International Nuclear Information System (INIS)

Rohm, H.H.

1987-01-01

Advanced technology offers significant potential benefit to the nuclear industry. Improvements can be anticipated in plant performance, reliability, and overall plant safety as well as reduced life cycle costs. Utilizing artificial intelligence and expert systems, robotics, advanced instruments and controls, and modularization technologies can enhance plant operations and provide new insights and perspectives to plant risk and thus focus resources to areas of importance. Plant reliability, operability, availability, accident interdiction and limitation, and plant recovery are expected to improve. However, utilizing these technologies is not an automatic process. In addition to the actual costs associated with developing and implementing the technologies, operator training and acceptance represents a potential significant problem. Traditional plant operators have little or no experience with computer technology. There has already been some difficulty getting nuclear plant operators to accept and use the new technologies that have been implemented to accept and use the new technologies that have been implemented thus far

Evaluating the Impact of Radio Frequency Identification Retained Surgical Instruments Tracking on Patient Safety: Literature Review.

Science.gov (United States)

Schnock, Kumiko O; Biggs, Bonnie; Fladger, Anne; Bates, David W; Rozenblum, Ronen

2017-02-22

Retained surgical instruments (RSI) are one of the most serious preventable complications in operating room settings, potentially leading to profound adverse effects for patients, as well as costly legal and financial consequences for hospitals. Safety measures to eliminate RSIs have been widely adopted in the United States and abroad, but despite widespread efforts, medical errors with RSI have not been eliminated. Through a systematic review of recent studies, we aimed to identify the impact of radio frequency identification (RFID) technology on reducing RSI errors and improving patient safety. A literature search on the effects of RFID technology on RSI error reduction was conducted in PubMed and CINAHL (2000-2016). Relevant articles were selected and reviewed by 4 researchers. After the literature search, 385 articles were identified and the full texts of the 88 articles were assessed for eligibility. Of these, 5 articles were included to evaluate the benefits and drawbacks of using RFID for preventing RSI-related errors. The use of RFID resulted in rapid detection of RSI through body tissue with high accuracy rates, reducing risk of counting errors and improving workflow. Based on the existing literature, RFID technology seems to have the potential to substantially improve patient safety by reducing RSI errors, although the body of evidence is currently limited. Better designed research studies are needed to get a clear understanding of this domain and to find new opportunities to use this technology and improve patient safety.

Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

International Nuclear Information System (INIS)

You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

2004-10-01

The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

Advanced nuclear power systems: Design, technology, safety and strategies for their deployment

International Nuclear Information System (INIS)

1994-01-01

The objectives of the symposium were to provide high level decision makers with an overview of the discussion concerning the need for nuclear power and salient features of advanced nuclear power systems; a forum for discussing the design objectives and safety approaches for such systems and the views of regulatory bodies; a forum for identifying barriers to the deployment of these systems and for reviewing strategies to overcome these barriers; and a forum for reviewing options for international cooperation in the development and deployment of such systems. Refs, figs and tabs

Safety-related instrumentation and control systems for nuclear power plants

International Nuclear Information System (INIS)

1984-01-01

This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety but are not safety systems. The Guide is intended to expand paragraphs 3.1, 3.2 and 3.3 of the Code of Practice on Design for Safety of Nuclear Power Plants (IAEA Safety Series No.50-C-D) in the area of I and C systems important to safety and refers to them as safety-related I and C systems. It also gives guidance and enumerates requirements for multiplexing and the use of the digital computers employed in this area

Report of the Advanced Neutron Source (ANS) safety workshop, Knoxville, Tennessee, October 25--26, 1988

International Nuclear Information System (INIS)

Buchanan, J.R.; Dumont, J.N.; Kendrick, C.M.; Row, T.H.; Thompson, P.B.; West, C.D.; Marchaterre, J.F.; Muhlheim, M.D.; McBee, M.R.

1988-12-01

On October 25--26, 1988, about 60 people took part in an Advanced Neutron Source (ANS) Safety Workshop, organized in cooperation with the Oak Ridge Operations (ORO) Office of the Department of Energy (DOE) and held in Knoxville, Tennessee. After a plenary session at which ANS Project staff presented status reports on the ANS design, research and development (R and D), and safety analysis efforts, the workshop broke into three working groups, each covering a different topic: Environmental and Waste Management, Applicable Regulatory Safety Criteria and Goals, and Reactor Concepts. Each group was asked to review the Project's approach to safety-related issues and to provide guidance on future reactor safety needs or directions for the Project. With the help of able chairmen, assisted by reporters and secretarial support, the working groups were extremely successful. Draft reports from each group were prepared before the workshop closed, and the major findings of each group were presented for review and discussion by the entire workshop attendance. This report contains the final version of the group reports, incorporating the results of the overall review by all the workshop participants

R and D in control, instrumentation, and electronic systems at CRNL

International Nuclear Information System (INIS)

Moeck, E.O.; Stirling, A.J.; Yan, G.

1980-10-01

Research and development in control, instrumentation, and electronics is part of the refinement of the CANDU power plant design and of the R and D work at the Chalk River Nuclear Laboratories. The programs presently being carried out in these areas are described in this report. Many of these projects originated in response to a potential for performance improvements in CANDU power plants and are aimed at extending plant capabilities and applying new technology and ideas to achieve better and less expensive systems. Some of these projects such as the development of in-reactor pressure transducers for fuel testing are to be completed in a matter of months. Others, for example heavy water vapor loss instruments, will have gone through the laboratory development phase within a year or two. More sophisicated developments such as multivariable controllers, computer-based safety systems, and advanced neutron detectors will continue over 4 or 5 years. This time scale also applies to the design and implementation of new control and safety systems for the NRU research reactor. Still others will require more than 5 years to be fully operational. New systems for the detection and location of failed fuel and a total information network with distributed intelligence, to be applied to future power stations, fall into that category. (auth)

Safety evaluation of advance street name signs

Science.gov (United States)

2009-06-01

The Federal Highway Administration (FHWA) organized a pooled fund study of 26 States to evaluate low-cost safety strategies as part of its strategic highway safety effort. The objective of the pooled fund study was to estimate the safety effectivenes...

Safety aspects of the US advanced LMR [liquid metal reactor] design

International Nuclear Information System (INIS)

Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.

1989-01-01

The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Quantitative assessment of probability of failing safely for the safety instrumented system using reliability block diagram method

International Nuclear Information System (INIS)

Jin, Jianghong; Pang, Lei; Zhao, Shoutang; Hu, Bin

2015-01-01

Highlights: • Models of PFS for SIS were established by using the reliability block diagram. • The more accurate calculation of PFS for SIS can be acquired by using SL. • Degraded operation of complex SIS does not affect the availability of SIS. • The safe undetected failure is the largest contribution to the PFS of SIS. - Abstract: The spurious trip of safety instrumented system (SIS) brings great economic losses to production. How to ensure the safety instrumented system is reliable and available has been put on the schedule. But the existing models on spurious trip rate (STR) or probability of failing safely (PFS) are too simplified and not accurate, in-depth studies of availability to obtain more accurate PFS for SIS are required. Based on the analysis of factors that influence the PFS for the SIS, using reliability block diagram method (RBD), the quantitative study of PFS for the SIS is carried out, and gives some application examples. The results show that, the common cause failure will increase the PFS; degraded operation does not affect the availability of the SIS; if the equipment was tested and repaired one by one, the unavailability of the SIS can be ignored; the corresponding occurrence time of independent safe undetected failure should be the system lifecycle (SL) rather than the proof test interval and the independent safe undetected failure is the largest contribution to the PFS for the SIS

Optimal Design of Safety Instrumented Systems for Pressure Control of Methanol Separation Columns in the Bisphenol a Manufacturing Process

Directory of Open Access Journals (Sweden)

In-Bok Lee

2016-12-01

Full Text Available A bisphenol A production plant possesses considerable potential risks in the top of the methanol separation column, as pressurized acetone, methanol, and water are processed at an elevated temperature, especially in the event of an abnormal pressure increase due to a sudden power outage. This study assesses the potential risks in the methanol separation column through hazard and operability assessments and evaluates the damages in the case of fire and explosion accident scenarios. The study chooses three leakage scenarios: a 5-mm puncture on the methanol separation column, a 50-mm diameter fracture of a discharge pipe and a catastrophic rupture, and, simulated using Phast (Ver. 6.531, the concentration distribution of scattered methanol, thermal radiation distribution of fires, and overpressure distribution of vapor cloud explosions. Implementation of a safety-instrumented system equipped with two-out-of-three voting as a safety measure can detect overpressure at the top of the column and shut down the main control valve and the emergency shutoff valve simultaneously. By applying a safety integrity level of three, the maximal release volume of the safety relief valve can be reduced and, therefore, the design capacity of the flare stack can also be reduced. Such integration will lead to improved safety at a reduced cost.

Advanced accumulator for PWR

International Nuclear Information System (INIS)

Ichimura, Taiki; Chikahata, Hideyuki

1997-01-01

Advanced accumulators have been incorporated into the APWR design in order to simplify the safety system configuration and to improve reliability. The advanced accumulators refill the reactor vessel with a large discharge flow rate in a large LOCA, then switch to a small flow rate to continue safety injection for core reflooding. The functions of the conventional accumulator and the low head safety injection pump are integrated into this advanced accumulator. Injection performance tests simulating LOCA conditions and visualization tests for new designs have been carried out. This paper describes the APWR ECCS configuration, the advanced accumulator design and some of the injection performance and visualization test results. It was verified that the flow resistance of the advanced accumulator is independent of the model scale. The similarity law and performance data of the advanced accumulator for applying APWR was established. (author)

Technology Roadmap on Instrumentation, Control, and Human-Machine Interface to Support DOE Advanced Nuclear Energy Programs

International Nuclear Information System (INIS)

Donald D Dudenhoeffer; Burce P Hallbert

2007-01-01

Instrumentation, Controls, and Human-Machine Interface (ICHMI) technologies are essential to ensuring delivery and effective operation of optimized advanced Generation IV (Gen IV) nuclear energy systems. In 1996, the Watts Bar I nuclear power plant in Tennessee was the last U.S. nuclear power plant to go on line. It was, in fact, built based on pre-1990 technology. Since this last U.S. nuclear power plant was designed, there have been major advances in the field of ICHMI systems. Computer technology employed in other industries has advanced dramatically, and computing systems are now replaced every few years as they become functionally obsolete. Functional obsolescence occurs when newer, more functional technology replaces or supersedes an existing technology, even though an existing technology may well be in working order. Although ICHMI architectures are comprised of much of the same technology, they have not been updated nearly as often in the nuclear power industry. For example, some newer Personal Digital Assistants (PDAs) or handheld computers may, in fact, have more functionality than the 1996 computer control system at the Watts Bar I plant. This illustrates the need to transition and upgrade current nuclear power plant ICHMI technologies

Account of requirements for modernization in VPBER-600 enhanced safety reactor instrumentation and control system development

International Nuclear Information System (INIS)

Shashkin, S.L.; Pobedonostsev, A.B.; Drumov, V.V.; Chudin, A.G.

1993-01-01

Nuclear power plant (NPP) with VPBER-600 reactor is a station of new generation. The specified term of reactor plant operation is 60 years and taking into account that the proposed term of starting the first power unit is on the turn of centuries one can definitely state that for Russia conditions VPBER-600 is a plant of 21 century. Such far removed term for NPP now in the stage of development as it can seem does not put the problems of modernization as first order tasks. But open-quotes...who does not think about future lives in the past.close quotes It is that the NPP instrumentation and control (I ampersand C) systems are in the most degree subjected to the influence of factors which favor their modifications. These factors can be arbitrarily divided into two groups: (1) inner factors, i.e. changes (failures, aging, etc) in I ampersand C components as well as changes dictated by technological reasons (change of equipment composition, control algorithms, operation modes); (2) outer factors, i.e. intensive development of information technologies and rapid improvement of electronic components. This presentation addresses the problem of modernization of the safety instrumentation for this next generation facility, and the research effort it will entail. The system is designed to allow for modernization, and the relatively easy adoption of new instrumentation and technology as it becomes available

76 FR 70953 - Pipeline Safety: Safety of Gas Transmission Pipelines

Science.gov (United States)

2011-11-16

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part 192 [Docket ID PHMSA-2011-0023] RIN 2137-AE72 Pipeline Safety: Safety of Gas Transmission Pipelines AGENCY: Pipeline and Hazardous Materials Safety Administration (PHMSA); DOT. ACTION: Advance notice of...

Safeguards instrumentation: past, present, future

International Nuclear Information System (INIS)

Higinbotham, W.A.

1982-01-01

Instruments are essential for accounting, for surveillance and for protection of nuclear materials. The development and application of such instrumentation is reviewed, with special attention to international safeguards applications. Active and passive nondestructive assay techniques are some 25 years of age. The important advances have been in learning how to use them effectively for specific applications, accompanied by major advances in radiation detectors, electronics, and, more recently, in mini-computers. The progress in seals has been disappointingly slow. Surveillance cameras have been widely used for many applications other than safeguards. The revolution in TV technology will have important implications. More sophisticated containment/surveillance equipment is being developed but has yet to be exploited. On the basis of this history, some expectations for instrumentation in the near future are presented

Instrumentation for Nuclear Applications

International Nuclear Information System (INIS)

1998-01-01

The objective of this project was to develop and coordinate nuclear instrumentation standards with resulting economies for the nuclear and radiation fields. There was particular emphasis on coordination and management of the Nuclear Instrument Module (NIM) System, U.S. activity involving the CAMAC international standard dataway system, the FASTBUS modular high-speed data acquisition and control system and processing and management of national nuclear instrumentation and detector standards, as well as a modest amount of assistance and consultation services to the Pollutant Characterization and Safety Research Division of the Office of Health and Environmental Research. The principal accomplishments were the development and maintenance of the NIM instrumentation system that is the predominant instrumentation system in the nuclear and radiation fields worldwide, the CAMAC digital interface system in coordination with the ESONE Committee of European Laboratories, the FASTBUS high-speed system and numerous national and international nuclear instrumentation standards

Instrumentation of steam cycle HTR's up to 900 MWe

International Nuclear Information System (INIS)

Leithner, D.E.; Winkenbach, B.

1982-06-01

Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

Advance of the National Program of Radiological Protection and Safety for medical diagnostic with X-rays

International Nuclear Information System (INIS)

Verdejo S, M.

1999-01-01

The National Program of Radiological Protection and Safety for medical diagnostic with X-ray (Programa Nacional de Proteccion y Seguridad Radiologica para diagnostico medico con rayos X) was initiated in the General Direction of Environmental Health (Direccion General de Salud Ambiental) in 1995. Task coordinated with different dependences of the Public Sector in collaboration between the Secretary of Health (Secretaria de Salud), the National Commission of Nuclear Safety and Safeguards (Comision Nacional de Seguridad Nuclear y Salvaguardias) and, the National Institute of Nuclear Research (Instituto Nacional de Investigaciones Nucleares). The surveillance to the fulfilment of the standardization in matter of Radiological Protection and Safety in the medical diagnostic with X-rays has been obtained for an important advance in the Public sector and it has been arousing interest in the Private sector. (Author)

Advanced ESPI-based medical instruments for otolaryngology

Science.gov (United States)

Castracane, James; Conerty, M.; Cacace, Anthony T.; Gardner, Glendon M.; Miller, Mitchell B.; Parnes, Steven M.

1993-05-01

Optical fibers have long been used for visual inspection inside the human body for medical diagnoses and treatment. By making use of sophisticated optical interferometric and ultra- small imaging techniques, combined with automated image processing, it is possible to extract significantly increased information for more accurate medical diagnoses. With support from NIH under the SBIR program, we have been developing a range of such instruments. One of these supported by the NIDCD is capable of providing detailed spatial information on the vibratory response of the tympanic membrane (TM). This instrument involves the examination of the TM by means of high speed electronic speckle pattern interferometry (ESPI). This provides a real time view of the vibration patterns of the TM for clinical diagnosis. This Interferometric Otoscope consists of mode conserving fiber optics, miniature diode lasers and high speed solid state detector arrays. We present the current status of the research including holography and ESPI of TM models and excised temporal bone preparations. A second instrument, also developed with support from NIDCD, is for application to the larynx. This system is also ESPI based but will incorporate features for direct vocal cord (VC) examination. By careful examination of the vibratory response of the VC during phonation, the characteristics of the mucosal wave may be examined. Adynamic regions of the cords can signal the start of lesions or cysts. Results of surgery can be evaluated in a quantitative manner. The design of a clinical prototype and preliminary electro-optic experiments on excised larynges and VC models will be presented.

Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

International Nuclear Information System (INIS)

St Leger-Barter, G.

1980-11-01

This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

Technical basis for instrumentation and control design improvements in WWER-440/230 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-01-01

Instrumentation and control (I and C) has been recognized as an area which requires substantial improvements in WWER NNPs, particularly for model 230 plants. Under contract with the IAEA the Spanish company Empresarios Agrupados (EA) developed a basic document proposing a technical basis for improvements related to the following most significant aspects of I and C: criteria for safety classification; remote shutdown panel; I and C support to operation and control room design; instrumentation set point margins; accident monitoring instrumentation. This publication is derived from the original report of EA which was circulated by the IAEA for review by staff members and experts from various Member States. It was finally agreed upon at a Consultants' meeting convened by the IAEA in Vienna in May 1994 with the participation of experts from France, Germany and Spain. The guidance expressed in this report is based on the IAEA/NUSS standards, safety guides and practices, and on regulations in use in various Member States. It is proposed as a way of carrying out the necessary studies to improve safety by upgrading the vital part of instrumentation an control in WWER-440 model 230 nuclear power plants. 28 refs, 3 figs

U.S. Department of Energy instrumentation and controls technology research for advanced small modular reactors

International Nuclear Information System (INIS)

Wood, Richard Thomas

2013-01-01

Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD and D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD and D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors. (author)

Probabilistic safety assessment for digital instrumentation and control systems in nuclear power plants - a review

International Nuclear Information System (INIS)

Lu, L.; Jiang, J.

2003-01-01

Deregulation in electricity market has created a great deal of challenges for nuclear power industries [1]. To stay competitive, Nuclear Power Plants (NPPs) will have to find ways to reduce their operational costs and to improve the plant safety. Instrumentation and Control (I and C) systems play an important role in this regard. Thus, new methodologies need to be developed to manage the operation of I and C systems more economically without jeopardizing the overall plant safety. Probabilistic Safety Assessment (PSA) technique is one of the promising methods to deal with such an issue, because PSA analyzes various system operational issues from a probabilistic sense, rather than a worst-case approach. However, there are several limitations when PSA is applied to I and C systems directly. A possible solution to this problem can be found by incorporating PSA with several other approaches. To better understand the issues involved, an attempt has been made in this paper to carry out a literature survey on this and related subject, particularly the effort will be made on: 1) the development of digital I and C systems in NPP, 2) PSA and its potential benefits and limitations, and 3) applications of PSA in various aspects of I and C systems including the resource allocation, the determination of surveillance testing strategies and the design of I and C systems. Finally, some solutions to overcome the aforementioned obstacles when applying PSA in I and C systems are also examined critically. (author)

Traceability of radiation protection instruments

Science.gov (United States)

Hino, Y.; Kurosawa, T.

2007-08-01

Radiation protection instruments are used in daily measurement of dose and activities in workplaces and environments for safety management. The requirements for calibration certificates with traceability are increasing for these instruments to ensure the consistency and reliabilities of the measurement results. The present traceability scheme of radiation protection instruments for dose and activity measurements is described with related IEC/ISO requirements. Some examples of desirable future calibration systems with recent new technologies are also discussed to establish the traceability with reasonable costs and reliabilities.

How could intelligent safety transport systems enhance safety ?

NARCIS (Netherlands)

Wiethoff, M. Heijer, T. & Bekiaris, E.

2017-01-01

In Europe, many deaths and injured each years are the cost of today's road traffic. Therefore, it is wise to look for possible solutions for enhancing traffic safety. Some Advanced Driver Assistance Systems (ADAS) are expected to increase safety, but they may also evoke new safety hazards. Only

Technical evaluation of the electrical, instrumentation, and control design aspects of the override of containment purge valve isolation and other engineered safety feature signals for the Fort Calhoun Nuclear Power Plant

International Nuclear Information System (INIS)

Hackett, D.B.

1980-01-01

This report documents the technical evaluation of the electrical, instrumentation, and control design aspects of the override of containment purge valve isolation and other engineered safety feature signals for the Fort Calhoun nuclear power plant. The review criteria are based on IEEE Std-279-1971 requirements for the safety signals to all purge and ventilation isolation valves. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

The impact of the instrumentation and control systems in the safety of a nuclear plant: a general vision; El impacto de los sistemas de instrumentacion y control en la seguridad de una planta nuclear: una vision general

Energy Technology Data Exchange (ETDEWEB)

Celis del Angel, L.; Rivero, T., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

One of the fundamental components so much for the sure operation, like in emergency cases or accident are the equipment s and instrumentation and control systems. The nuclear industry has had some accidents where the instrumentation and control have played and important part: a wrong design, instrumentation lack, faulty systems of safety, etc. At the present time the necessity to modernize the instrumentation and control in a nuclear power plant is before the challenge of finding innovative forms to improve the competitiveness and readiness, reducing operation costs without put ing in risk the safety and reliability of the nuclear power plant. Most of the nuclear power plants require actualizing their instrumentation and control systems, here the digital systems represent a great alternative, improving the performance and the safety, increasing the readiness and reducing the maintenance s. However they require of strict tests that allow assuring their application in critical systems. It is also necessary, the development of modernization programs that allow the programmed substitution of the systems without affecting the readiness of the nuclear power plants. During this whole modernization process will be necessary to put special attention in the cyber-safety because the attacks every time they are more elaborated. Therefore will be necessary to go toward the modernization of the instrumentation and control with the challenge of making without detriment some in the safety of the normal operation and with response reliability in emergency conditions or accident that which represents an effort that should not be postponed in the case of the nuclear power plant of Laguna Verde. (Author)

Smart antennas for nuclear instruments

International Nuclear Information System (INIS)

Jain, Ranjan Bala; Singhi, B.M.

2005-01-01

The advances in the field of computer and communications are leading to the development of smart embedded nuclear instruments. These instruments have highly sophisticated signal-processing algorithms based on FPGA and ASICS, provisions of present day connectivity and user interfaces. The developments in the connectivity, standards and bus technologies have made possible to access these instruments on LAN and WAN with suitable reliability and security. To get rid of wires i.e. in order to access these instruments, without wires at any place, wireless technology has evolved and become integral part of day-to-day activities. The environment monitoring can be done remotely, if smart antennas are incorporated on these instruments

Environmental tests of a digital safety channel: An investigation of stress-related vulnerabilities of computer-based safety system

International Nuclear Information System (INIS)

Korsah, K.; Wilson, T.L.; Wood, R.; Tanaka, T.

1997-01-01

This article presents the results of environmental stress tests performed on an experimental digital safety channel (EDSC) assembled at the Oak Ridge National Laboratory as part of the Qualification of Advanced Instrumentation and Controls Systems Research program, which was sponsored by the US Nuclear Regulatory Commission. The program is expected to provide recommendations for environmental qualification of digital safety systems. The purpose of the study was to investigate potential vulnerabilities of distributed computer systems used in safety applications when subjected to environmental stressors. The EDSC assembled for the tests employs technologies and digital subsystems representative of those proposed for use in advanced light-water reactors or as retrofits in existing plants. Subsystems include computers, electrical and optical serial communication links, fiber-optic network links, analog-to-digital and digital-to-analog converters, and multiplexers. The EDSC was subjected to selected stressors that are a potential risk to digital equipment in a mild environment. The selected stressors were electromagnetic and radiofrequency interferences (EMI-RFI), temperature, humidity, and smoke exposure. The stressors were applied at levels of intensity considerably higher than the safety channel is likely to experience in a normal nuclear power plant environment. Ranges of stress were selected at a sufficiently high level to induce errors so that failure modes that are characteristic of the technologies employed could be identified. On the basis of the incidence of functional errors observed during testing, EMI-RFI, smoke exposure, and high temperature coupled with high relative humidity, in that order, were found to have the greatest impact of the stressors investigated. The most prevalent stressor-induced upsets, as well as the most severe, were found to occur during the EMI-RFI tests

Efficacy and safety evaluation of icotinib in patients with advanced non-small cell lung cancer.

Science.gov (United States)

Gu, Aiqin; Shi, Chunlei; Xiong, Liwen; Chu, Tianqing; Pei, Jun; Han, Baohui

2013-02-01

To evaluate the efficacy and safety of icotinib hydrochloride in patients with advanced non-small cell lung cancer (NSCLC). A total of 89 patients with stage IIIB or IV NSCLC received icotinib at a dose of 125 mg administered 3 times a day. Icotinib treatment was continued until disease progression or development of unacceptable toxicity. A total of 89 patients were assessable. In patients treated with icotinib, the overall response rate (RR) was 36.0% (32/89), and the disease control rate (DCR) was 69.7% (62/89). RR and DCR were significantly improved in patients with adenocarcinoma versus non-adenocarcinoma (Picotinib hydrochloride in the treatment of advanced NSCLC is efficacious and safe, and its toxic effects are tolerable.

Outline of the advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

Hucik, S.A.; Imaoka, T.; Minematsu, A.; Takashima, Y.

1986-01-01

The fundamental design of the Advanced Boiling Water Reactor (ABWR) was completed in December 1985. This design represents the next generation of Boiling Water Reactors (BWR) to be introduced into commercial operation in the 1990s. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technologies and many new improvements based on an extensive accumulation of world-wide experience through design, construction and operation of BWRs. The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. The ABWR objectives were specific improvements such as operating and safety margins, enhanced availability and capacity factor, and reduced occupational exposure while at the same time achieving significant cost reduction in both capital and operating costs. The ABWR is characterized by an improved NSSS including ten internal recirculation pumps, fine motion electric-hydraulic control rod drives, optimized safety and auxiliary systems, advanced control and instrumentation systems, improved turbine-generator with moisture/separator reheater with plant output increased to 1350 MWe, and an integrated reinforced concrete containment vessel and compact Reactor and Turbine Building design. The turbine system also included improvements in the Turbine-Generator, feedwater/heater system, and condensate treatment systems. The radwaste system was also optimized taking advantage of the plant design improvements and advances in radwaste technology. The ABWR is a truly optimal design which utilizes advanced technologies, capabilities, performance improvements, and yet provides an economic advantage. (author)

Safety; Avertissement

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

Westinghouse AP600 advanced nuclear plant design

International Nuclear Information System (INIS)

Gangloff, W.

1999-01-01

As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

Recent advances of rearing cabinet instrumentation and control system for insect stock culture

Science.gov (United States)

Hermawan, Wawan; Kasmara, Hikmat; Melanie, Panatarani, Camellia; Joni, I. Made

2017-01-01

Helicoverpa armigera (Hubner) is one of a serious pest of horticulture in Indonesia. Helicoverpa armigera Nuclear Polyhedrovirus (HaNPV) has attracted interest for many researchers as a pest control for larvae of this species. Currently, we investigating the agrochemical formulations of HaNPV by introducing nanotechnology. Thus it is required an acceptable efficiency of insect stock cultures equipped with advance instruments to resolve the difficulties on insect stock seasons dependency. In addition, it is important to improve the insect survival with the aid of artificial natural environment and gain high insect production. This paper reports the rearing cabinet used as preparation of stock culture includes air-conditioning system, lighting, i.e. day and night control, and the main principles on recent technical and procedural advances apparatus of the system. The rearing system was moveable, designed and build by allowing air-conditioned cabinet for rearing insects, air motion and distribution as well as temperature and humidity being precisely controlled. The air was heated, humidified, and dehumidified respectively using a heater and ultrasonic nebulizer as actuators. Temperature and humidity can be controlled at any desired levels from room temperature (20°C) to 40 ± 1°C and from 0 to 80% RH with an accuracy of ±3% R.H. It is concluded that the recent design has acceptable performance based on the defined requirement for insect rearing and storage.

Digital control application for the advanced boiling water reactor

International Nuclear Information System (INIS)

Fennern, L.E.; Pearson, T.; Wills, H.D.; Swire Rhodes, L.; Pearson, R.L.

1986-01-01

The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

Nuclear power plants - Instrumentation and control systems important for safety - Classification (International Electrotechnical Commission Standard Publication 1226:1993)

International Nuclear Information System (INIS)

Stefanik, J.

1996-01-01

This international standard established a method of classification of the information and command functions for nuclear power plants, and the I and C and equipment that provide those functions, into categories that designate the importance for safety of the functions, and the associated systems and equipment. The resulting classification then determines relevant design criteria. The design criteria are the measures of quality by which the adequacy of each functions, and the associated systems and equipment in relation to its importance to plant safety is ensured. In this standard, the criteria are those of functionality, reliability, performance, environmental durability and quality assurance. This standard is applicable to all the information and command functions, and the instrumentation and control systems and equipment that provide those functions. The functions, systems and equipment under consideration provide automated protection, closed or open loop control, and information to the operating staff. They keep the NPP conditions inside the safe operating envelope and provide automatic actions, or enable manual actions, that mitigate accidents or prevent or minimize radioactive releases to the site or wider environment. The functions, and the associated systems and equipment that fulfill these roles safeguard the health and safety of the NPP operators and the public. This standard complements, and does not replace or supersede, the Safety Guides and Codes of Practice published by the International Atomic Energy Agency

An open-label extension study to evaluate the safety of ropinirole prolonged release in Chinese patients with advanced Parkinson's disease.

Science.gov (United States)

Zhang, Zhenxin; Wang, Jian; Zhang, Xiaoying; Chen, Shengdi; Wang, Zhenfu; Zhang, Baorong; Liu, Chunfeng; Qu, Qiumin; Cheng, Yan; Zhu, Rongxuan; Li, Jie; Hu, Jingqiu; Cai, Meng

2015-04-01

This open-label extension (OLE) study evaluated the safety profile of ropinirole prolonged release (PR) administered for 24 weeks as adjunctive to levodopa in Chinese patients with advanced Parkinson's disease (PD). This was a 24 week, flexible-dose, OLE study in Chinese patients with advanced PD who received 24 week treatment with ropinirole PR in the preceding double-blind (DB), phase III study (NCT01154166) and had no break in receiving study treatment while switching from the DB study to the OLE study. In the OLE study, patients received ropinirole PR once daily, starting with 2 mg/d and increasing up to 8 mg/d at week 4 (2 mg increment/week); if tolerable, the dose could be further increased in 4 mg increments up to 24 mg/d according to clinical judgment. There were no efficacy assessments. Safety assessments included monitoring adverse events (AEs), neurological examination, Gambling Symptom Assessment Scale questionnaire, liver chemistry, and laboratory tests. Of the 295 enrolled patients, 282 completed the study. The most common reason for withdrawal was AEs (n = 9, 3.1%). The mean duration to ropinirole PR treatment was 173.1 days and an overall median daily dose of ropinirole was 8 mg (range: 2-24 mg). Overall, 114 (38.6%) patients experienced on-treatment AEs; the most frequent reported AEs ( ≥ 2%) were dyskinesia (6.1%), dizziness (4.1%), nausea (3.4%), hallucinations (3.4%), somnolence (2.7%) and decreased weight (2.4%). Sixty-eight patients (23.1%) experienced treatment-related AEs. Six patients experienced serious AEs (SAEs), of which hallucination was determined to be a treatment-related SAE. There were no other significant safety findings. No new safety signals for ropinirole were identified. The safety profile of ropinirole was consistent with the preceding DB study and also with the established safety profile for ropinirole. Results support the long-term use of ropinirole PR as an adjunctive to levodopa in Chinese patients

Reliability analysis and computation of computer-based safety instrumentation and control used in German nuclear power plant. Final report; Zuverlaessigkeitsuntersuchung und -berechnung rechnerbasierter Sicherheitsleittechnik zum Einsatz in deutschen Kernkraftwerken. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Ding, Yongjian [Hochschule Magdeburg-Stendal, Magdeburg (Germany). Inst. fuer Elektrotechnik; Krause, Ulrich [Magdeburg Univ. (Germany). Inst. fuer Apparate- und Umwelttechnik; Gu, Chunlei

2014-08-21

The trend of technological advancement in the field of safety instrumentation and control (I and C) leads to increasingly frequent use of computer-based (digital) control systems which consisting of distributed, connected bus communications computers and their functionalities are freely programmable by qualified software. The advantages of the new I and C system over the old I and C system with hard-wired technology are e.g. in the higher flexibility, cost-effective procurement of spare parts, higher hardware reliability (through higher integration density, intelligent self-monitoring mechanisms, etc.). On the other hand, skeptics see the new technology with the computer-based I and C a higher potential by influences of common cause failures (CCF), and the easier manipulation by sabotage (IT Security). In this joint research project funded by the Federal Ministry for Economical Affaires and Energy (BMWi) (2011-2014, FJZ 1501405) the Otto-von-Guericke-University Magdeburg and Magdeburg-Stendal University of Applied Sciences are therefore trying to develop suitable methods for the demonstration of the reliability of the new instrumentation and control systems with the focus on the investigation of CCF. This expertise of both houses shall be extended to this area and a scientific contribution to the sound reliability judgments of the digital safety I and C in domestic and foreign nuclear power plants. First, the state of science and technology will be worked out through the study of national and international standards in the field of functional safety of electrical and I and C systems and accompanying literature. On the basis of the existing nuclear Standards the deterministic requirements on the structure of the new digital I and C system will be determined. The possible methods of reliability modeling will be analyzed and compared. A suitable method called multi class binomial failure rate (MCFBR) which was successfully used in safety valve applications will be

High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

International Nuclear Information System (INIS)

1983-01-01

The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

Development of the advanced CANDU technology

International Nuclear Information System (INIS)

Suk, Soo Dong; Min, Byung Joo; Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R.

1997-07-01

The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs

Development of the advanced CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y H; Lee, S Y; Choi, J H; Lee, B C; Kim, S N; Jo, C H; Paik, J S; On, M R; Park, H S; Kim, S R [Korea Electric Power Co., Taejon (Korea, Republic of)

1997-07-01

The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

Fieldcrest Cannon, Inc. Advanced Technical Preparation. Statistical Process Control (SPC). Safety Section: Modules 1-3. Instructor Book.

Science.gov (United States)

Averitt, Sallie D.

These three modules, which were developed for use by instructors in a manufacturing firm's advanced technical preparation program, contain the materials required to present the safety section of the plant's adult-oriented, job-specific competency-based training program. The 3 modules contain 12 lessons on the following topics: lockout/tagout…

Advanced Instrumentation for Positron Emission Tomography [PET

Science.gov (United States)

Derenzo, S. E.; Budinger, T. F.

1985-04-01

This paper summarizes the physical processes and medical science goals that underlay modern instrumentation design for Positron Emission Tomography. The paper discusses design factors such as detector material, crystalphototube coupling, shielding geometry, sampling motion, electronics design, time-of-flight, and the interrelationships with quantitative accuracy, spatial resolution, temporal resolution, maximum data rates, and cost.

Seismic instrumentation

International Nuclear Information System (INIS)

1984-06-01

RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The aim of this RFS is to define the type, location and operating conditions for seismic instrumentation needed to determine promptly the seismic response of nuclear power plants features important to safety to permit comparison of such response with that used as the design basis

Safety climate in Swiss hospital units: Swiss version of the Safety Climate Survey

Science.gov (United States)

Gehring, Katrin; Mascherek, Anna C.; Bezzola, Paula

2015-01-01

Abstract Rationale, aims and objectives Safety climate measurements are a broadly used element of improvement initiatives. In order to provide a sound and easy‐to‐administer instrument for the use in Swiss hospitals, we translated the Safety Climate Survey into German and French. Methods After translating the Safety Climate Survey into French and German, a cross‐sectional survey study was conducted with health care professionals (HCPs) in operating room (OR) teams and on OR‐related wards in 10 Swiss hospitals. Validity of the instrument was examined by means of Cronbach's alpha and missing rates of the single items. Item‐descriptive statistics group differences and percentage of ‘problematic responses’ (PPR) were calculated. Results 3153 HCPs completed the survey (response rate: 63.4%). 1308 individuals were excluded from the analyses because of a profession other than doctor or nurse or invalid answers (n = 1845; nurses = 1321, doctors = 523). Internal consistency of the translated Safety Climate Survey was good (Cronbach's alpha G erman = 0.86; Cronbach's alpha F rench = 0.84). Missing rates at item level were rather low (0.23–4.3%). We found significant group differences in safety climate values regarding profession, managerial function, work area and time spent in direct patient care. At item level, 14 out of 21 items showed a PPR higher than 10%. Conclusions Results indicate that the French and German translations of the Safety Climate Survey might be a useful measurement instrument for safety climate in Swiss hospital units. Analyses at item level allow for differentiating facets of safety climate into more positive and critical safety climate aspects. PMID:25656302

Meteorological instrumentation

International Nuclear Information System (INIS)

1982-06-01

RFS or ''Regles Fondamentales de Surete'' (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety , while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the ''Service Central de Surete des Installations Nucleaires'' or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to specify the meteorological instrumentation required at the site of each nuclear power plant equipped with at least one pressurized water reactor

Instrumentation and control of future sodium cooled fast reactors - Design improvements

International Nuclear Information System (INIS)

Madhusoodanan, K.; Sakthivel, M.; Chellapandi, P.

2013-06-01

India's fast reactor program started with the 40 MWt Fast Breeder Test Reactor. 500 MWe Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam. Safety of PFBR is enhanced by improved design features of I and C system. Since the design of Instrumentation and control (I and C) of PFBR, considerable improvements in terms of advancement in technology and indigenization has taken place. Further improvements in I and C is proposed for solving many of the difficulties faced during the design and construction phases of PFBR. Design improvements proposed are covered in this paper which will make the implementation and maintenance of I and C of future SFRs easier. (authors)

Book of extended synopses. International symposium on advanced nuclear power systems. Design, technology, safety and strategies for their deployment

International Nuclear Information System (INIS)

1993-01-01

During the symposium the design, technology, safety and strategy for the development of advanced nuclear power systems were discussed. 20 papers were presented at the symposium. A separate abstract was prepared for each of these papers. Refs

Seismic instrumentation for nuclear power plants

International Nuclear Information System (INIS)

Senne Junior, M.

1983-01-01

A seismic instrumentation system used in Nuclear Power Plants to monitor the design parameters of systems, structures and components, needed to provide safety to those Plants, against the action of earthquakes is described. The instrumentation described is based on the nuclear standards in force. The minimum amount of sensors and other components used, as well as their general localization, is indicated. The operation of the instrumentation system as a whole and the handling of the recovered data are dealt with accordingly. The various devices used are not covered in detail, except for the accelerometer, which is the seismic instrumentation basic component. (Author) [pt

Development of a tool of probabilistic safety analysis for advanced nuclear reactors

International Nuclear Information System (INIS)

Hidalgo H, F.E.; Fran N, P.

2007-01-01

It is developing a tool to explain in a simple way in that it consists the Probabilistic Safety Analysis (APS) and at the same time to facilitate the comparison among the different designs of advanced nuclear reactors starting from their safety systems. This tool for teaching contemplates all the workspaces in an APS, but it is deepened only in what is the development of accident sequences and systems models. At the moment its have incorporated three types of advanced reactors, ABWR, ESBWR, and the HTGR and they are compared among if and with a BWR like that of Laguna Verde. This tool is carried out in Visual Basic code because it is a platform that can be used in any Windows atmosphere and for their easy programming. The system includes a tree of events developed for this purpose for a research HTGR built in Japan (HTTR) to have a point of comparison of the same one with other reactors of previous generations. It is that in the fourth generation reactors the measure of frequency of core damage doesn't make the same sense that for reactors of previous generations, which is due to its passive safety systems and its design type of the fuel, that which makes indispensable the development of another type of risk measure. The tree of events is presented for the initiator event 'the rupture of the main pipe' that causes the depressurization of the HTTR reactor. In this article it was concluded that it is necessary to evaluate the accident until reaching to the liberation of fission products that one knows in APS like an APS study level 1 and level 2 together. The final states developed starting from the possible phenomena that happen in these scenarios are presented. For this, its are considered flaws of all the mitigation systems that intervene in this accident. The tree of events developed for this work and the definition of the final states contributes to the development of as carrying out an APS for fourth generation reactors, with the purpose of developing an APS

Advances in Airborne and Ground Geophysical Methods for Uranium Exploration

International Nuclear Information System (INIS)

2013-01-01

One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property.' The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. According to current estimates, after 2050 available resources and cumulative uranium demand are expected to be balanced. Recent annual uranium production constitutes 65-75% of annual nuclear energy uranium requirements worldwide. With typical timeframes of between 15 and 40 years from the commencement of uranium exploration activities until uranium production, together with increasingly favourable uranium prices, many countries have seen a rise in uranium exploration activities since 2005. This experience and assessment of global energy needs support continued uranium exploration

Seismic instrumentation for nuclear power plants

International Nuclear Information System (INIS)

Senne Junior, M.

1983-07-01

A seismic instrumentation system used in Nuclear Power Plants to monitor the design parameters of systems, structures and components, needed to provide safety to those plants, against the action of earth quarks is described. The instrumentation is based on the nuclear standards and other components used, as well as their general localization is indicated. The operation of the instrumentation system as a whole and the handling of the recovered data are dealt with accordingly. The accelerometer is described in detail. (Author) [pt

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

Advanced reactor experimental facilities

International Nuclear Information System (INIS)

Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

2010-01-01

For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

Probabilistic safety assessment

International Nuclear Information System (INIS)

Hoertner, H.; Schuetz, B.

1982-09-01

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

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

International Nuclear Information System (INIS)

2004-01-01

More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

International Nuclear Information System (INIS)

2004-01-01

More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

Advanced Safeguards Technology Road-map for the Global Nuclear Energy Partnership

International Nuclear Information System (INIS)

Miller, M.C.; Tobin, S.; Smith, L.E.; Ehinger, M.; Dougan, A.; Cipiti, B.; Bakel, A.; Bean, R.

2008-01-01

Strengthening the nonproliferation regime, including advanced safeguards, is a cornerstone of the Global Nuclear Energy Partnership (GNEP). To meet these challenges, the Safeguards Campaign was formed, whose mission is to provide research and technology development for the foundation of next generation safeguards systems for implementation in U.S. GNEP facilities. The Safeguards Campaign works closely with the Nuclear Nonproliferation and International Security department (NA-24) of NNSA (National Nuclear Safety Administration) to ensure that technology developed for domestic safeguards applications are optimum with respect to international safeguards use. A major milestone of the program this year has been the development of the advanced safeguards technology road-map. This paper will broadly describe the road-map, which provides a path to next generation safeguards systems including advanced instrumentation; process monitoring; data integration, protection, and analysis; and system level evaluation and knowledge extraction for real time applications. (authors)

ACR-700 advanced technologies

International Nuclear Information System (INIS)

Tapping, R.L.; Turner, C.W.; Yu, S.K.W.; Olmstead, R.; Speranzini, R.A.

2004-01-01

A successful advanced reactor plant will have optimized economics including reduced operating and maintenance costs, improved performance, and enhanced safety. Incorporating improvements based on advanced technologies ensures cost, safety and operational competitiveness of the ACR-700. These advanced technologies include modern configuration management; construction technologies; operational technology for the control centre and information systems for plant monitoring and analysis. This paper summarizes the advanced technologies used to achieve construction and operational improvements to enhance plant economic competitiveness, advances in the operational technology used for reactor control, and presents the development of the Smart CANDU suite of tools and its application to existing operating reactors and to the ACR-700. (author)

The Fiber Optic System for the Advanced Topographic Laser Altimeter System (ATLAS) Instrument

Science.gov (United States)

Ott, Melanie N.; Thomes, Joe; Onuma, Eleanya; Switzer, Robert; Chuska, Richard; Blair, Diana; Frese, Erich; Matyseck, Marc

2016-01-01

The Advanced Topographic Laser Altimeter System (ATLAS) Instrument has been in integration and testing over the past 18 months in preparation for the Ice, Cloud and Land Elevation Satellite - 2 (ICESat-2) Mission, scheduled to launch in 2017. ICESat-2 is the follow on to ICESat which launched in 2003 and operated until 2009. ATLAS will measure the elevation of ice sheets, glaciers and sea ice or the "cryosphere" (as well as terrain) to provide data for assessing the earth's global climate changes. Where ICESat's instrument, the Geo-Science Laser Altimeter (GLAS) used a single beam measured with a 70 m spot on the ground and a distance between spots of 170 m, ATLAS will measure a spot size of 10 m with a spacing of 70 cm using six beams to measure terrain height changes as small as 4 mm. The ATLAS pulsed transmission system consists of two lasers operating at 532 nm with transmitter optics for beam steering, a diffractive optical element that splits the signal into 6 separate beams, receivers for start pulse detection and a wavelength tracking system. The optical receiver telescope system consists of optics that focus all six beams into optical fibers that feed a filter system that transmits the signal via fiber assemblies to the detectors. Also included on the instrument is a system that calibrates the alignment of the transmitted pulses to the receiver optics for precise signal capture. The larger electro optical subsystems for transmission, calibration, and signal receive, stay aligned and transmitting sufficiently due to the optical fiber system that links them together. The robust design of the fiber optic system, consisting of a variety of multi fiber arrays and simplex assemblies with multiple fiber core sizes and types, will enable the system to maintain consistent critical alignments for the entire life of the mission. Some of the development approaches used to meet the challenging optical system requirements for ATLAS are discussed here.

The fiber optic system for the Advanced Topographic Laser Altimeter System (ATLAS) instrument.

Science.gov (United States)

Ott, Melanie N; Thomes, Joe; Onuma, Eleanya; Switzer, Robert; Chuska, Richard; Blair, Diana; Frese, Erich; Matyseck, Marc

2016-08-28

The Advanced Topographic Laser Altimeter System (ATLAS) Instrument has been in integration and testing over the past 18 months in preparation for the Ice, Cloud and Land Elevation Satellite - 2 (ICESat-2) Mission, scheduled to launch in 2017. ICESat-2 is the follow on to ICESat which launched in 2003 and operated until 2009. ATLAS will measure the elevation of ice sheets, glaciers and sea ice or the "cryosphere" (as well as terrain) to provide data for assessing the earth's global climate changes. Where ICESat's instrument, the Geo-Science Laser Altimeter (GLAS) used a single beam measured with a 70 m spot on the ground and a distance between spots of 170 m, ATLAS will measure a spot size of 10 m with a spacing of 70 cm using six beams to measure terrain height changes as small as 4 mm.[1] The ATLAS pulsed transmission system consists of two lasers operating at 532 nm with transmitter optics for beam steering, a diffractive optical element that splits the signal into 6 separate beams, receivers for start pulse detection and a wavelength tracking system. The optical receiver telescope system consists of optics that focus all six beams into optical fibers that feed a filter system that transmits the signal via fiber assemblies to the detectors. Also included on the instrument is a system that calibrates the alignment of the transmitted pulses to the receiver optics for precise signal capture. The larger electro optical subsystems for transmission, calibration, and signal receive, stay aligned and transmitting sufficiently due to the optical fiber system that links them together. The robust design of the fiber optic system, consisting of a variety of multi fiber arrays and simplex assemblies with multiple fiber core sizes and types, will enable the system to maintain consistent critical alignments for the entire life of the mission. Some of the development approaches used to meet the challenging optical system requirements for ATLAS are discussed here.

Assessing hospitals' clinical risk management: Development of a monitoring instrument

Directory of Open Access Journals (Sweden)

Pfeiffer Yvonne

2010-12-01

Full Text Available Abstract Background Clinical risk management (CRM plays a crucial role in enabling hospitals to identify, contain, and manage risks related to patient safety. So far, no instruments are available to measure and monitor the level of implementation of CRM. Therefore, our objective was to develop an instrument for assessing CRM in hospitals. Methods The instrument was developed based on a literature review, which identified key elements of CRM. These elements were then discussed with a panel of patient safety experts. A theoretical model was used to describe the level to which CRM elements have been implemented within the organization. Interviews with CRM practitioners and a pilot evaluation were conducted to revise the instrument. The first nationwide application of the instrument (138 participating Swiss hospitals was complemented by in-depth interviews with 25 CRM practitioners in selected hospitals, for validation purposes. Results The monitoring instrument consists of 28 main questions organized in three sections: 1 Implementation and organizational integration of CRM, 2 Strategic objectives and operational implementation of CRM at hospital level, and 3 Overview of CRM in different services. The instrument is available in four languages (English, German, French, and Italian. It allows hospitals to gather comprehensive and systematic data on their CRM practice and to identify areas for further improvement. Conclusions We have developed an instrument for assessing development stages of CRM in hospitals that should be feasible for a continuous monitoring of developments in this important area of patient safety.

Assessing hospitals' clinical risk management: Development of a monitoring instrument.

Science.gov (United States)

Briner, Matthias; Kessler, Oliver; Pfeiffer, Yvonne; Wehner, Theo; Manser, Tanja

2010-12-13

Clinical risk management (CRM) plays a crucial role in enabling hospitals to identify, contain, and manage risks related to patient safety. So far, no instruments are available to measure and monitor the level of implementation of CRM. Therefore, our objective was to develop an instrument for assessing CRM in hospitals. The instrument was developed based on a literature review, which identified key elements of CRM. These elements were then discussed with a panel of patient safety experts. A theoretical model was used to describe the level to which CRM elements have been implemented within the organization. Interviews with CRM practitioners and a pilot evaluation were conducted to revise the instrument. The first nationwide application of the instrument (138 participating Swiss hospitals) was complemented by in-depth interviews with 25 CRM practitioners in selected hospitals, for validation purposes. The monitoring instrument consists of 28 main questions organized in three sections: 1) Implementation and organizational integration of CRM, 2) Strategic objectives and operational implementation of CRM at hospital level, and 3) Overview of CRM in different services. The instrument is available in four languages (English, German, French, and Italian). It allows hospitals to gather comprehensive and systematic data on their CRM practice and to identify areas for further improvement. We have developed an instrument for assessing development stages of CRM in hospitals that should be feasible for a continuous monitoring of developments in this important area of patient safety.

Analytical chemistry instrumentation

International Nuclear Information System (INIS)

Laing, W.R.

1986-01-01

In nine sections, 48 chapters cover 1) analytical chemistry and the environment 2) environmental radiochemistry 3) automated instrumentation 4) advances in analytical mass spectrometry 5) fourier transform spectroscopy 6) analytical chemistry of plutonium 7) nuclear analytical chemistry 8) chemometrics and 9) nuclear fuel technology

Multilevel model of safety climate for furniture industries.

Science.gov (United States)

Rodrigues, Matilde A; Arezes, Pedro M; Leão, Celina P

2015-01-01

Furniture companies can analyze their safety status using quantitative measures. However, the data needed are not always available and the number of accidents is under-reported. Safety climate scales may be an alternative. However, there are no validated Portuguese scales that account for the specific attributes of the furniture sector. The current study aims to develop and validate an instrument that uses a multilevel structure to measure the safety climate of the Portuguese furniture industry. The Safety Climate in Wood Industries (SCWI) model was developed and applied to the safety climate analysis using three different scales: organizational, group and individual. A multilevel exploratory factor analysis was performed to analyze the factorial structure. The studied companies' safety conditions were also analyzed. Different factorial structures were found between and within levels. In general, the results show the presence of a group-level safety climate. The scores of safety climates are directly and positively related to companies' safety conditions; the organizational scale is the one that best reflects the actual safety conditions. The SCWI instrument allows for the identification of different safety climates in groups that comprise the same furniture company and it seems to reflect those groups' safety conditions. The study also demonstrates the need for a multilevel analysis of the studied instrument.

Current status of environmental, health, and safety issues of electrochemical capacitors for advanced vehicle applications

Energy Technology Data Exchange (ETDEWEB)

Vimmerstedt, L J; Hammel, C J

1997-04-01

Electrochemical capacitors are a candidate for traction power assists in hybrid electric vehicles (HEVs). Other advanced automotive applications, while not the primary focus of current development efforts, are also possible. These include load leveling high-energy batteries, power conditioning electronics, electrically hated catalysts, electric power steering, and engine starter power. Higher power and longer cycle life are expected for electrochemical capacitors than for batteries. Evaluation of environmental, health, and safety (EH and S) issues of electrochemical capacitors is an essential part of the development and commercialization of electrochemical capacitors for advanced vehicles. This report provides an initial EH and S assessment. This report presents electrochemical capacitor electrochemistry, materials selection, intrinsic material hazards, mitigation of those hazards, environmental requirements, pollution control options, and shipping requirements. Most of the information available for this assessment pertains to commercial devices intended for application outside the advanced vehicle market and to experiment or prototype devices. Electrochemical capacitors for power assists in HEVs are not produced commercially now. Therefore, materials for advanced vehicle electrochemical capacitors may change, and so would the corresponding EH and S issues. Although changes are possible, this report describes issues for likely electrochemical capacitor designs.

Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

Modeling safety instrumented systems with MooN voting architectures addressing system reconfiguration for testing

International Nuclear Information System (INIS)

Torres-Echeverria, A.C.; Martorell, S.; Thompson, H.A.

2011-01-01

This paper addresses the modeling of probability of dangerous failure on demand and spurious trip rate of safety instrumented systems that include MooN voting redundancies in their architecture. MooN systems are a special case of k-out-of-n systems. The first part of the article is devoted to the development of a time-dependent probability of dangerous failure on demand model with capability of handling MooN systems. The model is able to model explicitly common cause failure and diagnostic coverage, as well as different test frequencies and strategies. It includes quantification of both detected and undetected failures, and puts emphasis on the quantification of common cause failure to the system probability of dangerous failure on demand as an additional component. In order to be able to accommodate changes in testing strategies, special treatment is devoted to the analysis of system reconfiguration (including common cause failure) during test of one of its components, what is then included in the model. Another model for spurious trip rate is also analyzed and extended under the same methodology in order to empower it with similar capabilities. These two models are powerful enough, but at the same time simple, to be suitable for handling of dependability measures in multi-objective optimization of both system design and test strategies for safety instrumented systems. The level of modeling detail considered permits compliance with the requirements of the standard IEC 61508. The two models are applied to brief case studies to demonstrate their effectiveness. The results obtained demonstrated that the first model is adequate to quantify time-dependent PFD of MooN systems during different system states (i.e. full operation, test and repair) and different MooN configurations, which values are averaged to obtain the PFD avg . Also, it was demonstrated that the second model is adequate to quantify STR including spurious trips induced by internal component failure and

Multi-objective optimization of design and testing of safety instrumented systems with MooN voting architectures using a genetic algorithm

International Nuclear Information System (INIS)

Torres-Echeverría, A.C.; Martorell, S.; Thompson, H.A.

2012-01-01

This paper presents the optimization of design and test policies of safety instrumented systems using MooN voting redundancies by a multi-objective genetic algorithm. The objectives to optimize are the Average Probability of Dangerous Failure on Demand, which represents the system safety integrity, the Spurious Trip Rate and the Lifecycle Cost. In this way safety, reliability and cost are included. This is done by using novel models of time-dependent probability of failure on demand and spurious trip rate, recently published by the authors. These models are capable of delivering the level of modeling detail required by the standard IEC 61508. Modeling includes common cause failure and diagnostic coverage. The Probability of Failure on Demand model also permits to quantify results with changing testing strategies. The optimization is performed using the multi-objective Genetic Algorithm NSGA-II. This allows weighting of the trade-offs between the three objectives and, thus, implementation of safety systems that keep a good balance between safety, reliability and cost. The complete methodology is applied to two separate case studies, one for optimization of system design with redundancy allocation and component selection and another for optimization of testing policies. Both optimization cases are performed for both systems with MooN redundancies and systems with only parallel redundancies. Their results are compared, demonstrating how introducing MooN architectures presents a significant improvement for the optimization process.

CSNI collective statement on support facilities for existing and advanced reactors. The function of OECD/Nea joint projects Nea committee on the safety of nuclear installations (CSNI)

International Nuclear Information System (INIS)

2008-01-01

The NEA Committee on the Safety of Nuclear Installations (CSNI) has recently completed a study on the availability and utilisation of facilities supporting safety studies for current and advanced nuclear power reactors. The study showed that significant steps had been undertaken in the past several years in support of safety test facilities, mainly by conducting multinational joint projects centered on the capability of unique test facilities worldwide. Given the positive experience of the safety research projects, it has been recommended that efforts be made to prioritize technical issues associated with advanced (Generation IV) reactor designs and to develop options on how to efficiently obtain the necessary data through internationally co-ordinated research, preparing a gradual extension of safety research beyond the needs set by currently operating reactors. This statement constitutes a reference for future CSNI activities and for safety authorities, R and D centres and industry for internationally co-ordinated research initiatives in the nuclear safety research area. (author)

Simulation tools for detector and instrument design

DEFF Research Database (Denmark)

Kanaki, Kalliopi; Kittelmann, Thomas; Cai, Xiao Xiao

2018-01-01

The high performance requirements at the European Spallation Source have been driving the technological advances on the neutron detector front. Now more than ever is it important to optimize the design of detectors and instruments, to fully exploit the ESS source brilliance. Most of the simulation...... a powerful set of tools to tailor the detector and instrument design to the instrument application....

Nuclear instrumentation for radiation measurement

International Nuclear Information System (INIS)

Madan, V.K.

2012-01-01

Nuclear radiation cannot be detected by human senses. Nuclear detectors and associated electronics facilitate detection and measurement of different types of radiation like alpha particles, beta particles, gamma radiation, and detection of neutrons. Nuclear instrumentation has evolved greatly since the discovery of radioactivity. There has been tremendous advancement in detector technology, electronics, computer technology, and development of efficient algorithms and methods for spectral processing to extract precisely qualitative and quantitative information of the radiation. Various types of detectors and nuclear instruments are presently available and are used for different applications. This paper describes nuclear radiation, its detection and measurement and associated electronics, spectral information extraction, and advances in these fields. The paper also describes challenges in this field

The Frequency of Nonmotor Symptoms among Advanced Parkinson Patients May Depend on Instrument Used for Assessment

Directory of Open Access Journals (Sweden)

Nelson Hwynn

2011-01-01

Full Text Available Background. Nonmotor symptoms (NMS of Parkinson's disease (PD may be more debilitating than motor symptoms. The purpose of this study was to determine the frequency and corecognition of NMS among our advanced PD cohort (patients considered for deep brain stimulation (DBS and caregivers. Methods. NMS-Questionnaire (NMS-Q, a self-administered screening questionnaire, and NMS Assessment-Scale (NMS-S, a clinician-administered scale, were administered to PD patients and caregivers. Results. We enrolled 33 PD patients (23 males, 10 females and caregivers. The most frequent NMS among patients using NMS-Q were gastrointestinal (87.9%, sleep (84.9%, and urinary (72.7%, while the most frequent symptoms using NMS-S were sleep (90.9%, gastrointestinal (75.8%, and mood (75.8%. Patient/caregiver scoring correlations for NMS-Q and NMS-S were 0.670 (<0.0001 and 0.527 (=0.0016, respectively. Conclusion The frequency of NMS among advanced PD patients and correlation between patients and caregivers varied with the instrument used. The overall correlation between patient and caregiver was greater with NMS-Q than NMS-S.

Advanced korean industrial safety and health policy with risk assessment.

Science.gov (United States)

Kwon, Hyuckmyun; Cho, Jae Hyun; Moon, Il; Choi, Jaewook; Park, Dooyong; Lee, Youngsoon

2010-09-01

This article describes a systematic roadmap master plan for advanced industrial safety and health policy in Korea, with an emphasis on. Since Korean industries had first emergence of industrial safety and health policy in 1953, enormous efforts have been made on upgrading the relevant laws in order to reflect real situation of industrial work environment in accordance with rapid changes of Korean and global business over three decades. Nevertheless, current policy has major defects; too much techniques-based articles, diverged contents in less organization, combined enforcement and punishments and finally enforcing regulations full of commands and control. These deficiencies have make it difficult to accommodate changes of social, industrial and employment environment in customized fashion. The approach to the solution must be generic at the level of paradigm-shift rather than local modifications and enhancement. The basic idea is to establish a new system integrated with a risk assessment scheme, which encourages employers to apply to their work environment under comprehensive responsibility. The risk assessment scheme is designed to enable to inspect employers' compliances afterwards. A project comprises four yearly phases based on applying zones; initially designating and operating a specified risk zone, gradually expanding the special zones during a period of 3 years (2010-2012) and the final zone expanded to entire nation. In each phase, the intermediate version of the system is updated through a process of precise and unbiased validation in terms of its operability, feasibility and sustainability with building relevant infrastructures as needed.

Analysis approach for common cause failure on non-safety digital control system

Energy Technology Data Exchange (ETDEWEB)

Kim, Yun Goo; Oh, Eungse [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2014-05-15

The effects of common cause failure (CCF) on safety digital instrumentation and control (I and C) system had been considered in defense in depth and diversity coping analysis with safety analysis method. For the non-safety system, single failure had been considered for safety analysis. IEEE Std. 603-1991, Clause 5.6.3.1(2), 'Isolation' states that no credible failure on the non-safety side of an isolation device shall prevent any portion of a safety system from meeting its minimum performance requirements during and following any design basis event requiring that safety function. The software CCF is one of the credible failure on the non-safety side. In advanced digital I and C system, same hardware component is used for different control system and the defect in manufacture or common external event can generate CCF. Moreover, the non-safety I and C system uses complex software for its various function and software quality assurance for the development process is less severe than safety software for the cost effective design. Therefore the potential defects in software cannot be ignored and the effect of software CCF on non-safety I and C system is needed to be evaluated. This paper proposes the general process and considerations for the analysis of CCF on non-safety I and C system.